ML20311A628: Difference between revisions

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{{#Wiki_filter:1Q/2000 Inspection Findings - Callaway                                                                                                    Page 1 of 12 Callaway Initiating Events Significance:        Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
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During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor
 
1Q/2000 Inspection Findings - Callaway                                                                                                  Page 2 of 12 because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
 
1Q/2000 Inspection Findings - Callaway                                                                                                  Page 3 of 12 Significance:        Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Significance:        Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause
 
1Q/2000 Inspection Findings - Callaway                                                                                                  Page 4 of 12 analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event.
Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program.
Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
 
1Q/2000 Inspection Findings - Callaway                                                                                                Page 5 of 12 Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections
 
1Q/2000 Inspection Findings - Callaway                                                                                                    Page 6 of 12 in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Barrier Integrity Significance:        Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B
 
1Q/2000 Inspection Findings - Callaway                                                                                                Page 7 of 12 required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
 
1Q/2000 Inspection Findings - Callaway                                                                                                Page 8 of 12 Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable
 
1Q/2000 Inspection Findings - Callaway                                                                                                Page 9 of 12 performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white
 
1Q/2000 Inspection Findings - Callaway                                                                                                Page 10 of 12 findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded.
The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Public Radiation Safety Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 -
Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
 
1Q/2000 Inspection Findings - Callaway                                                                                                Page 11 of 12 Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection Miscellaneous Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective.
However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
 
1Q/2000 Inspection Findings - Callaway                                                                                              Page 12 of 12 Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Last modified : April 01, 2002
 
2Q/2000 Inspection Findings - Callaway                                                                                                    Page 1 of 12 Callaway Initiating Events Significance:        Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary
 
2Q/2000 Inspection Findings - Callaway                                                                                                    Page 2 of 12 feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing
 
2Q/2000 Inspection Findings - Callaway                                                                                                  Page 3 of 12 Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
 
2Q/2000 Inspection Findings - Callaway                                                                                                  Page 4 of 12 Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Significance:        Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On
 
2Q/2000 Inspection Findings - Callaway                                                                                                Page 5 of 12 December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event.
Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the
 
2Q/2000 Inspection Findings - Callaway                                                                                                  Page 6 of 12 ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program.
Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Barrier Integrity Significance:          Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Significance:          Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to
 
2Q/2000 Inspection Findings - Callaway                                                                                                Page 7 of 12 correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance:        Jun 30, 2001
 
2Q/2000 Inspection Findings - Callaway                                                                                                Page 8 of 12 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
 
2Q/2000 Inspection Findings - Callaway                                                                                                Page 9 of 12 Inspection Report# : 2001008(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on
 
2Q/2000 Inspection Findings - Callaway                                                                                                Page 10 of 12 January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded.
The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Public Radiation Safety Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 -
Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2000 Inspection Findings - Callaway                                                                                                Page 11 of 12 Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection Miscellaneous Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective.
However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation
 
2Q/2000 Inspection Findings - Callaway                                                                                              Page 12 of 12 Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Last modified : April 01, 2002
 
3Q/2000 Inspection Findings - Callaway                                                                                                    Page 1 of 12 Callaway Initiating Events Significance:          Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program.
Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
 
3Q/2000 Inspection Findings - Callaway                                                                                                Page 2 of 12 Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections
 
3Q/2000 Inspection Findings - Callaway                                                                                                    Page 3 of 12 in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not
 
3Q/2000 Inspection Findings - Callaway                                                                                                  Page 4 of 12 have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC
 
3Q/2000 Inspection Findings - Callaway                                                                                                  Page 5 of 12 Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Significance:          Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
 
3Q/2000 Inspection Findings - Callaway                                                                                                Page 6 of 12 This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event.
Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Barrier Integrity Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Significance:        Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to
 
3Q/2000 Inspection Findings - Callaway                                                                                                Page 7 of 12 correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC
 
3Q/2000 Inspection Findings - Callaway                                                                                                Page 8 of 12 identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded.
The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Significance:          Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance:          Aug 10, 2001
 
3Q/2000 Inspection Findings - Callaway                                                                                                Page 9 of 12 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
 
3Q/2000 Inspection Findings - Callaway                                                                                                Page 10 of 12 Inspection Report# : 2001008(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Public Radiation Safety Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 -
Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2000 Inspection Findings - Callaway                                                                                                Page 11 of 12 Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection Miscellaneous Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective.
However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation
 
3Q/2000 Inspection Findings - Callaway                                                                                              Page 12 of 12 Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Last modified : March 29, 2002
 
4Q/2000 Inspection Findings - Callaway                                                                                                    Page 1 of 12 Callaway Initiating Events Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Significance:        Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Mitigating Systems Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
 
4Q/2000 Inspection Findings - Callaway                                                                                                  Page 2 of 12 Inspection Report# : 2000015(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program.
Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
 
4Q/2000 Inspection Findings - Callaway                                                                                                    Page 3 of 12 Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment
 
4Q/2000 Inspection Findings - Callaway                                                                                                  Page 4 of 12 was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service
 
4Q/2000 Inspection Findings - Callaway                                                                                                  Page 5 of 12 A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
 
4Q/2000 Inspection Findings - Callaway                                                                                                  Page 6 of 12 Significance:        Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Significance:        Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event.
Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Barrier Integrity Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Significance:        Jan 10, 2001
 
4Q/2000 Inspection Findings - Callaway                                                                                                Page 7 of 12 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Emergency Preparedness Significance:          Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage
 
4Q/2000 Inspection Findings - Callaway                                                                                                Page 8 of 12 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded.
The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended
 
4Q/2000 Inspection Findings - Callaway                                                                                                Page 9 of 12 dose.
Inspection Report# : 2001005(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee
 
4Q/2000 Inspection Findings - Callaway                                                                                                Page 10 of 12 intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Public Radiation Safety Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 -
Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
 
4Q/2000 Inspection Findings - Callaway                                                                                                Page 11 of 12 Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection Miscellaneous Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
 
4Q/2000 Inspection Findings - Callaway                                                                                            Page 12 of 12 Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective.
However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Last modified : March 28, 2002
 
1Q/2001 Inspection Findings - Callaway                                                                                                    Page 1 of 12 Callaway Initiating Events Significance:        Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:        Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event.
Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation
 
1Q/2001 Inspection Findings - Callaway                                                                                                  Page 2 of 12 Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program.
Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation
 
1Q/2001 Inspection Findings - Callaway                                                                                                Page 3 of 12 consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-
 
1Q/2001 Inspection Findings - Callaway                                                                                                    Page 4 of 12 Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
 
1Q/2001 Inspection Findings - Callaway                                                                                                  Page 5 of 12 A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
 
1Q/2001 Inspection Findings - Callaway                                                                                                  Page 6 of 12 Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Barrier Integrity Significance:          Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Significance:          Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-
 
1Q/2001 Inspection Findings - Callaway                                                                                                Page 7 of 12 Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is
 
1Q/2001 Inspection Findings - Callaway                                                                                                Page 8 of 12 reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded.
The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Significance:          Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
 
1Q/2001 Inspection Findings - Callaway                                                                                                Page 9 of 12 Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation
 
1Q/2001 Inspection Findings - Callaway                                                                                                Page 10 of 12 of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Public Radiation Safety Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 -
Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC
 
1Q/2001 Inspection Findings - Callaway                                                                                                Page 11 of 12 Item Type: NCV NonCited Violation Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection Miscellaneous Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective.
However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC
 
1Q/2001 Inspection Findings - Callaway                                                                                              Page 12 of 12 Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : March 28, 2002
 
2Q/2001 Inspection Findings - Callaway                                                                                                    Page 1 of 12 Callaway Initiating Events Significance:        Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:        Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take
 
2Q/2001 Inspection Findings - Callaway                                                                                                  Page 2 of 12 corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Significance:          Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2001 Inspection Findings - Callaway                                                                                                  Page 3 of 12 Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event.
Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in
 
2Q/2001 Inspection Findings - Callaway                                                                                                Page 4 of 12 Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program.
Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C
 
2Q/2001 Inspection Findings - Callaway                                                                                                  Page 5 of 12 A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
 
2Q/2001 Inspection Findings - Callaway                                                                                                    Page 6 of 12 Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Barrier Integrity Significance:        Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation
 
2Q/2001 Inspection Findings - Callaway                                                                                                Page 7 of 12 consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
 
2Q/2001 Inspection Findings - Callaway                                                                                                Page 8 of 12 Significance:          Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding
 
2Q/2001 Inspection Findings - Callaway                                                                                                Page 9 of 12 Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded.
The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess
 
2Q/2001 Inspection Findings - Callaway                                                                                                Page 10 of 12 dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
Public Radiation Safety Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 -
Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee
 
2Q/2001 Inspection Findings - Callaway                                                                                              Page 11 of 12 Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Physical Protection Miscellaneous Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective.
However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2001 Inspection Findings - Callaway                                                                                              Page 12 of 12 Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Last modified : March 27, 2002
 
3Q/2001 Inspection Findings - Callaway                                                                                                    Page 1 of 12 Callaway Initiating Events Significance:          Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor
 
3Q/2001 Inspection Findings - Callaway                                                                                                  Page 2 of 12 because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:          Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Significance:          Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2001 Inspection Findings - Callaway                                                                                                  Page 3 of 12 Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event.
Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
 
3Q/2001 Inspection Findings - Callaway                                                                                                  Page 4 of 12 Significance:          Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program.
Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
 
3Q/2001 Inspection Findings - Callaway                                                                                                Page 5 of 12 Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections
 
3Q/2001 Inspection Findings - Callaway                                                                                                    Page 6 of 12 in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Barrier Integrity Significance:        Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B
 
3Q/2001 Inspection Findings - Callaway                                                                                                Page 7 of 12 required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
 
3Q/2001 Inspection Findings - Callaway                                                                                                Page 8 of 12 Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is
 
3Q/2001 Inspection Findings - Callaway                                                                                                Page 9 of 12 considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs
 
3Q/2001 Inspection Findings - Callaway                                                                                                Page 10 of 12 that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded.
The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Public Radiation Safety Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 -
Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
 
3Q/2001 Inspection Findings - Callaway                                                                                              Page 11 of 12 Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Physical Protection Miscellaneous Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective.
However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
 
3Q/2001 Inspection Findings - Callaway                                                                                              Page 12 of 12 Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Last modified : March 26, 2002
 
4Q/2001 Inspection Findings - Callaway                                                                                                    Page 1 of 11 Callaway Initiating Events Significance:        Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
 
4Q/2001 Inspection Findings - Callaway                                                                                                  Page 2 of 11 Significance:          Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
 
4Q/2001 Inspection Findings - Callaway                                                                                                  Page 3 of 11 Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Significance:        Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event.
Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
 
4Q/2001 Inspection Findings - Callaway                                                                                                  Page 4 of 11 Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program.
Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
Significance:          Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
 
4Q/2001 Inspection Findings - Callaway                                                                                                    Page 5 of 11 Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
 
4Q/2001 Inspection Findings - Callaway                                                                                                Page 6 of 11 Barrier Integrity Significance:        Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The
 
4Q/2001 Inspection Findings - Callaway                                                                                                Page 7 of 11 safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as
 
4Q/2001 Inspection Findings - Callaway                                                                                                Page 8 of 11 low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is
 
4Q/2001 Inspection Findings - Callaway                                                                                                Page 9 of 11 reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:        Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded.
The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Public Radiation Safety Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 -
Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2001 Inspection Findings - Callaway                                                                                                Page 10 of 11 Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection Miscellaneous Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective.
However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
 
4Q/2001 Inspection Findings - Callaway Page 11 of 11 Inspection Report# : 2000011(pdf)
Last modified : March 01, 2002
 
1Q/2002 Inspection Findings - Callaway                                                                                      Page 1 of 15 Callaway Initiating Events Significance:          Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process.
Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet. A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:          Mar 13, 2002 Identified By: NRC Item Type: NCV NonCited Violation Foreign object renders B Essential Service Water pump inoperable A noncited violation of Technical Specification 3.0.3 occurred five times during the time that the Essential Service Water pump was inoperable, three of which exceeded the one hour requirement for initiating actions identified in Technical Specification 3.0.3.
Specifically, on February 14, 2001, at 8:51 a.m., the licensee declared the ESW Pump B inoperable due to a tygon tube drain line becoming entwined around the pump impeller. At the same time, Containment Cooler C was out of service for planned maintenance.
This met the conditions for entry into TS 3.0.3. The licensee restored the containment cooler to service at 11:15 a.m., which was 2 hours and 32 minutes after when Technical Specification 3.0.3. should have been entered. Four other instances were identified where TS 3.0.3 should have been entered, two of the four times exceeded the one-hour action requirement identified in the TS. Due to the fact that the licensee was unaware that the ESW pump was inoperable from 2:15 p.m. on February 9 until 8:51a.m. on February 14, 2001, they had not realized that they had entered TS 3.0.3 several times. The finding was more than minor because it had an actual impact on safety in that one of the essential service water pumps was rendered inoperable for a duration greater than the allowed outage time while the plant was in a mode of operation that requires the ESW system to be operable. This finding was found to be of very low safety significance because the other train of Essential Service Water was always operable, and there was no actual emergency requiring the operation of the essential service water system. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200100515, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2002008(pdf)
Significance:          Mar 13, 2002 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2002 Inspection Findings - Callaway                                                                                    Page 2 of 15 Failure to promptly identify the need for and implement corrective action to address the degraded condition of the Auxiliary Feedwater System Train B During the independent review, the team determined that the licensee failed to promptly identify the need for and implement corrective action to address the flow anomaly condition of the auxiliary feedwater system Train B that existed between February 2000 and March 28, 2001, where the flow through the recirculation valve was below the required flow. The condition had a credible impact on safety since the flow anomaly had only been addressed from the standpoint of pump performance and operability and not system performance and required train function. However, since there was no actual loss of safety function and the system would have delivered the required minimum of 500 gpm to two steam generators when the function was required, the finding was considered to be of very low safety significance. Because of the very low safety significance and because the licensee included the item in their corrective action program by reopening Callaway Action Request 200000669 on March 1, 2002, this violation is being treated as a noncited violation (50-483/0208-01) in accordance with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2002008(pdf)
Significance: N/A Mar 13, 2002 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of the event that occurred between February 9 - 15, 2001, where one train of Essential Service Water had been rendered inoperable for approximately 132 hours. If a loss of offsite power had occurred while a train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their safety function.
The finding was previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/01-09. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes pertaining to the inoperable Essential Service Water pump and correctly identified the extent of the conditions for having one train of Essential Service Water inoperable for approximately 132 hours. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) personnel did not know that they needed to secure the drain hose because corrective action from a previous event did not preclude foreign material from entering the suction bay for the essential service water pump, (2) the drain hose was not adequately secured because there was no procedure for the job, (3) the drain hose was not adequately secured because important information that should have been covered during the pre-job brief was omitted, (4) personnel did not know that they needed to secure the drain hose because safety precautions and warnings were left out of the work package, (5) personnel that saw or were informed of the presence of a funnel without a drain hose did not have a questioning attitude, (6) the control room took over one hour to enter Technical Specification 3.0.3 after declaring "B" Essential Service Water system inoperable because personnel found the procedure difficult to use, and (7) the control room took over one hour to enter Technical Specification 3.0.3 after declaring "B" Essential Service Water system inoperable because training was not repeated enough times so that information could be learned and skills practiced. With regard to the extent of condition, the licensee found that the first five root causes identified extended throughout the plant for both installation of leakage control devices and foreign material exclusion. The licensee specified appropriate corrective actions to address the root causes and had implemented these actions by January, 2002. Because of the licensee's acceptable performance in addressing the inoperability of the "B" Essential Service Water system, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2002008(pdf)
Significance:        Feb 27, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to verify calculational methods.
Calculations for auxiliary feedwater pump net positive suction head did not account for nitrogen saturated water. The failure of calculational methods to verify the adequacy of net positive suction head requirements for the auxiliary feedwater pumps was a violation of 10 CFR Part 50, Appendix B, Criterion III. The failure to account for nitrogen saturated water in the net positive suction head calculation for the AFW pumps was more than minor because there was a credible impact on safety in that the available margin of net positive suction head was reduced by 11 feet. Using Phase 1 of the Significance Determination Process, the issue was determined to be of very low safety significance because adequate available net positive suction head remained after accounting for dissolved nitrogen. Therefore, the auxiliary feedwater pump would have remained available during an actual plant event. The finding was entered in the licensee's corrective action program as Callaway Action Report System Item CARS 200200485.
Inspection Report# : 2002007(pdf)
Significance:        Feb 27, 2002 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a significant condition adverse to quality.
Between January 1992 and January 31, 2002, several opportunities were missed to promptly identify and correct a significant
 
1Q/2002 Inspection Findings - Callaway                                                                                  Page 3 of 15 condition adverse to quality involving foreign material in the auxiliary feedwater system and condensate storage tank. The failure to promptly identify the degraded condition resulted in the failure of an auxiliary feedwater pump on December 3, 2001. In addition, between January 25 and 29, 2002, the identification of a significant condition adverse to quality involving the as-found condition of the degraded diaphragm seal was not reported to the appropriate levels of management. The multiple examples of missed opportunities to identify a significant condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and also represented a significant human performance cross cutting issue involving the timely recognition of degraded conditions. The finding had greater than minor significance because there was a credible impact on plant safety. Specifically, auxiliary feedwater Pump A failed to run when started by operations personnel during a plant shutdown. Had a plant event occurred, the potential existed for foam from the degraded condensate storage tank diaphragm to fail one or more auxiliary feedwater pumps. The failure of an auxiliary feedwater pump would have adversely affected the decay heat removal critical safety function. A Significance Determination Process Phase 3 analysis preliminarily determined that the issue had low to moderate safety significance (White).
This finding was entered in the licensee's corrective action program as Callaway Action Request System Item CARS 200107423.
Inspection Report# : 2002007(pdf)
Significance:        Feb 08, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate corrective action to address auxiliary feedwater system vibration.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that abnormally high vibration on both motor driven trains of the auxiliary feedwater system was corrected. During the past 12 years, the licensee had identified this condition five times. The licensee did not determine the actual cause of auxiliary feedwater piping vibration and consequently did not take appropriate corrective action. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had a credible impact on safety in that, if this vibration had occurred when auxiliary feedwater was needed, it could have affected operation of the system. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because the likelihood that the system would be operated in the condition that caused the abnormally high vibrations was low, nondestructive examinations revealed no piping degradation, and because no vibrations were observed on the turbine driven auxiliary feedwater train. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200200881, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001007(pdf)
Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected. Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C. The instructions failed to include guidance on how much oil to add to pump bearings following maintenance. Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action
 
1Q/2002 Inspection Findings - Callaway                                                                                      Page 4 of 15 program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored.
With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power. This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable.
Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001
 
1Q/2002 Inspection Findings - Callaway                                                                                      Page 5 of 15 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness. Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Significance:        May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate calculation of diesel loading.
Requirements in Procedure EDP-ZZ-04023, "Calculations", Revision 14, were not applied correctly to the diesel generator steady-state loading calculation contained in Callaway Drawing E-21005, "List of Loads Supplied by Emergency Diesel Generator,"
Revision 25. The drawing functioned as a calculation, but lacked the quality requirements for calculations required by this procedure.
The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the diesel generators retained adequate margin). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CAR 200203017), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy Inspection Report# : 2002004(pdf)
 
1Q/2002 Inspection Findings - Callaway                                                                                    Page 6 of 15 Significance:          May 24, 2002 Identified By: NRC Item Type: FIN Finding Incomplete and incorrect methods to evaluate degraded voltage conditions.
Two licensee calculations contained incomplete and incorrect methods of evaluating degraded voltage conditions. Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, did not consider the voltage requirements for non-motor loads in determining the degraded voltage relay setting. In addition, Calculation ZZ-214, "Motor Operated Valve Feeder Cable Voltage Drops," Addenda 1, Revision 2, for determining minimum voltage to motor-operated valves, did not consider the effect of motor starting currents in circuit elements upstream of the motor control centers. This finding, which did not involve a violation of NRC requirements, was of very low safety significance because the calculation errors did not result in an actual loss of safety function (the degraded voltage relay setpoint remained valid).
Inspection Report# : 2002004(pdf)
Significance:          May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to control design input for degraded voltage relay calculation.
Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, used to determine the degraded voltage relay dropout setting, referred to superseded calculations for important design inputs, and had not been updated to reflect plant configuration and loading changes. This was contrary to the requirement in Procedure EDP-ZZ-04023 that calculations be revised whenever a new or revised calculation (having an effect on the calculation) is issued. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the degraded voltage relay setpoint remained valid). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CARs 200203080 and 200203057), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy.
Inspection Report# : 2002004(pdf)
Significance:          Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event. Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:          Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train
 
1Q/2002 Inspection Findings - Callaway                                                                                    Page 7 of 15 remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:        Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles.
Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire. Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program. Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
 
1Q/2002 Inspection Findings - Callaway                                                                                    Page 8 of 15 Inspection Report# : 2000013(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar.
This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058).
The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period.
Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were
 
1Q/2002 Inspection Findings - Callaway                                                                                      Page 9 of 15 operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions.
Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Barrier Integrity Significance:        Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months.
The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B.
This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Emergency Preparedness
 
1Q/2002 Inspection Findings - Callaway                                                                                  Page 10 of 15 Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02.
Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field." Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact
 
1Q/2002 Inspection Findings - Callaway                                                                                Page 11 of 15 instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance.
Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Significance:        Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during
 
1Q/2002 Inspection Findings - Callaway                                                                                    Page 12 of 15 Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101 (b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable.
Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.) The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101 (b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable.
Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:          Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101 (b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable.
Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
 
1Q/2002 Inspection Findings - Callaway                                                                                    Page 13 of 15 Significance:          Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups. Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101 (b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable.
Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2001007(pdf)
Significance:          Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded. The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Public Radiation Safety Significance:          Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7 - Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 - Radioactive Material UN2982." Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
 
1Q/2002 Inspection Findings - Callaway                                                                                  Page 14 of 15 Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit. However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection Miscellaneous Significance: N/A Feb 27, 2002 Identified By: NRC Item Type: FIN Finding Deficiencies with implementation of corrective action and operability evaluation programs.
The team determined that several opportunities were missed to promptly identify and correct a risk significant condition adverse to quality involving the degraded condition of the condensate storage tank diaphragm seal. Quality assurance personnel were not actively involved in providing oversight of the event review team and root cause investigation processes. The event review team process did not ensure that statements were obtained from all personnel involved in the event. The corrective action program did not include guidance or expectations on the assignment of appropriate resources to review the highest classification of significant conditions adverse to quality. Minimal resources were initially assigned to the root cause investigation and may have contributed to the delay in identifying the degraded diaphragm seal. Based on interviews with the licensee's staff and a review of the corrective action program procedure, the team determined that licensed operators were only notified of equipment deficiencies if the individual discovering the condition believed there was an immediate impact on nuclear, plant, or personnel safety. Consequently, the potential existed for operability decisions to be made by non-licensed personnel. The operability evaluation program did not implement the guidance provided in NRC Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions."
Inspection Report# : 2002007(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation
 
1Q/2002 Inspection Findings - Callaway                                                                                    Page 15 of 15 Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities.
On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective. However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule.
This was a Severity Level IV violation of 10 CFR 50.65(a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Last modified : July 22, 2002
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 1 of 21 Callaway Initiating Events Significance:      Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:      Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet.
A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance:      Jun 25, 2002 Identified By: NRC Item Type: NCV NonCited Violation Unsecured fire door.
A noncited violation of Operating License Condition 2.C(5)(c) occurred when the licensee failed to take compensatory action when the 3-hour rated fire doors that separated the two trains of control room air conditioning were unlatched and not closed. This finding is more than minor because if a fire had occurred while the doors were unlatched and not closed, they could not perform their function of preventing a fire from spreading from one fire area to another fire area.
This finding affected the mitigating system cornerstone. This finding was evaluated using Appendix F of the reactor safety significance determination process and determined to be of very low safety significance because the combustible load for the area was low and because the fire detectors on each side of the doors were operable. This finding was entered into the licensee's corrective action system as Callaway Action Request System Number 200204041.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 2 of 21 Inspection Report# : 2002002(pdf)
Significance:      May 24, 2002 Identified By: NRC Item Type: FIN Finding Incomplete and incorrect methods to evaluate degraded voltage conditions.
Two licensee calculations contained incomplete and incorrect methods of evaluating degraded voltage conditions.
Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, did not consider the voltage requirements for non-motor loads in determining the degraded voltage relay setting. In addition, Calculation ZZ-214, "Motor Operated Valve Feeder Cable Voltage Drops," Addenda 1, Revision 2, for determining minimum voltage to motor-operated valves, did not consider the effect of motor starting currents in circuit elements upstream of the motor control centers. This finding, which did not involve a violation of NRC requirements, was of very low safety significance because the calculation errors did not result in an actual loss of safety function (the degraded voltage relay setpoint remained valid).
Inspection Report# : 2002004(pdf)
Significance:      May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to control design input for degraded voltage relay calculation.
Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, used to determine the degraded voltage relay dropout setting, referred to superseded calculations for important design inputs, and had not been updated to reflect plant configuration and loading changes. This was contrary to the requirement in Procedure EDP-ZZ-04023 that calculations be revised whenever a new or revised calculation (having an effect on the calculation) is issued. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the degraded voltage relay setpoint remained valid). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CARs 200203080 and 200203057), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy.
Inspection Report# : 2002004(pdf)
Significance:      May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate calculation of diesel loading.
Requirements in Procedure EDP-ZZ-04023, "Calculations", Revision 14, were not applied correctly to the diesel generator steady-state loading calculation contained in Callaway Drawing E-21005, "List of Loads Supplied by Emergency Diesel Generator," Revision 25. The drawing functioned as a calculation, but lacked the quality requirements for calculations required by this procedure. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the diesel generators retained adequate margin). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CAR 200203017), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy Inspection Report# : 2002004(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                      Page 3 of 21 Significance:      Apr 23, 2002 Identified By: Self Disclosing Item Type: FIN Finding Foreign material in condensate transfer system.
A leather weld rod pouch lodged inside the fill valve to the condensate storage tank could have adversely affected the auxiliary feedwater system if the pouch became dislodged while filling the tank. This finding is more than minor because the lack of foreign material exclusion controls could have resulted in the leather weld rod pouch entering the condensate storage tank and adversely affecting the auxiliary feedwater system. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance using the reactor safety significance determination process because no loss of safety function occurred and only one of three auxiliary feedwater pumps would have been affected. This finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200202678.
Inspection Report# : 2002002(pdf)
Significance:      Mar 13, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify the need for and implement corrective action to address the degraded condition of the Auxiliary Feedwater System Train B During the independent review, the team determined that the licensee failed to promptly identify the need for and implement corrective action to address the flow anomaly condition of the auxiliary feedwater system Train B that existed between February 2000 and March 28, 2001, where the flow through the recirculation valve was below the required flow. The condition had a credible impact on safety since the flow anomaly had only been addressed from the standpoint of pump performance and operability and not system performance and required train function. However, since there was no actual loss of safety function and the system would have delivered the required minimum of 500 gpm to two steam generators when the function was required, the finding was considered to be of very low safety significance. Because of the very low safety significance and because the licensee included the item in their corrective action program by reopening Callaway Action Request 200000669 on March 1, 2002, this violation is being treated as a noncited violation (50-483/0208-01) in accordance with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2002008(pdf)
Significance: N/A Mar 13, 2002 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of the event that occurred between February 9 - 15, 2001, where one train of Essential Service Water had been rendered inoperable for approximately 132 hours. If a loss of offsite power had occurred while a train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their safety function. The finding was previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/01-09. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes pertaining to the inoperable Essential Service Water pump and correctly identified the extent of the conditions for having one train of Essential Service Water inoperable for approximately 132 hours.
The licensee's evaluation identified the primary root causes of the performance issues to be: (1) personnel did not know that they needed to secure the drain hose because corrective action from a previous event did not preclude foreign material from entering the suction bay for the essential service water pump, (2) the drain hose was not adequately secured because there was no procedure for the job, (3) the drain hose was not adequately secured because important file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                          Page 4 of 21 information that should have been covered during the pre-job brief was omitted, (4) personnel did not know that they needed to secure the drain hose because safety precautions and warnings were left out of the work package, (5) personnel that saw or were informed of the presence of a funnel without a drain hose did not have a questioning attitude, (6) the control room took over one hour to enter Technical Specification 3.0.3 after declaring "B" Essential Service Water system inoperable because personnel found the procedure difficult to use, and (7) the control room took over one hour to enter Technical Specification 3.0.3 after declaring "B" Essential Service Water system inoperable because training was not repeated enough times so that information could be learned and skills practiced. With regard to the extent of condition, the licensee found that the first five root causes identified extended throughout the plant for both installation of leakage control devices and foreign material exclusion. The licensee specified appropriate corrective actions to address the root causes and had implemented these actions by January, 2002. Because of the licensee's acceptable performance in addressing the inoperability of the "B" Essential Service Water system, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2002008(pdf)
Significance:      Mar 13, 2002 Identified By: NRC Item Type: NCV NonCited Violation Foreign object renders B Essential Service Water pump inoperable A noncited violation of Technical Specification 3.0.3 occurred five times during the time that the Essential Service Water pump was inoperable, three of which exceeded the one hour requirement for initiating actions identified in Technical Specification 3.0.3. Specifically, on February 14, 2001, at 8:51 a.m., the licensee declared the ESW Pump B inoperable due to a tygon tube drain line becoming entwined around the pump impeller. At the same time, Containment Cooler C was out of service for planned maintenance. This met the conditions for entry into TS 3.0.3. The licensee restored the containment cooler to service at 11:15 a.m., which was 2 hours and 32 minutes after when Technical Specification 3.0.3. should have been entered. Four other instances were identified where TS 3.0.3 should have been entered, two of the four times exceeded the one-hour action requirement identified in the TS. Due to the fact that the licensee was unaware that the ESW pump was inoperable from 2:15 p.m. on February 9 until 8:51a.m. on February 14, 2001, they had not realized that they had entered TS 3.0.3 several times. The finding was more than minor because it had an actual impact on safety in that one of the essential service water pumps was rendered inoperable for a duration greater than the allowed outage time while the plant was in a mode of operation that requires the ESW system to be operable. This finding was found to be of very low safety significance because the other train of Essential Service Water was always operable, and there was no actual emergency requiring the operation of the essential service water system. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200100515, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2002008(pdf)
Significance:      Feb 27, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to verify calculational methods.
Calculations for auxiliary feedwater pump net positive suction head did not account for nitrogen saturated water. The failure of calculational methods to verify the adequacy of net positive suction head requirements for the auxiliary feedwater pumps was a violation of 10 CFR Part 50, Appendix B, Criterion III. The failure to account for nitrogen saturated water in the net positive suction head calculation for the AFW pumps was more than minor because there was a credible impact on safety in that the available margin of net positive suction head was reduced by 11 feet. Using file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 5 of 21 Phase 1 of the Significance Determination Process, the issue was determined to be of very low safety significance because adequate available net positive suction head remained after accounting for dissolved nitrogen. Therefore, the auxiliary feedwater pump would have remained available during an actual plant event. The finding was entered in the licensee's corrective action program as Callaway Action Report System Item CARS 200200485.
Inspection Report# : 2002007(pdf)
Significance:      Feb 27, 2002 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a significant condition adverse to quality.
Between January 1992 and January 31, 2002, several opportunities were missed to promptly identify and correct a significant condition adverse to quality involving foreign material in the auxiliary feedwater system and condensate storage tank. The failure to promptly identify the degraded condition resulted in the failure of an auxiliary feedwater pump on December 3, 2001. In addition, between January 25 and 29, 2002, the identification of a significant condition adverse to quality involving the as-found condition of the degraded diaphragm seal was not reported to the appropriate levels of management. The multiple examples of missed opportunities to identify a significant condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and also represented a significant human performance cross cutting issue involving the timely recognition of degraded conditions. The finding had greater than minor significance because there was a credible impact on plant safety. Specifically, auxiliary feedwater Pump A failed to run when started by operations personnel during a plant shutdown. Had a plant event occurred, the potential existed for foam from the degraded condensate storage tank diaphragm to fail one or more auxiliary feedwater pumps. The failure of an auxiliary feedwater pump would have adversely affected the decay heat removal critical safety function. A Significance Determination Process Phase 3 analysis preliminarily determined that the issue had low to moderate safety significance (White). This finding was entered in the licensee's corrective action program as Callaway Action Request System Item CARS 200107423.
Inspection Report# : 2002007(pdf)
Significance:      Feb 08, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate corrective action to address auxiliary feedwater system vibration.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that abnormally high vibration on both motor driven trains of the auxiliary feedwater system was corrected. During the past 12 years, the licensee had identified this condition five times. The licensee did not determine the actual cause of auxiliary feedwater piping vibration and consequently did not take appropriate corrective action. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had a credible impact on safety in that, if this vibration had occurred when auxiliary feedwater was needed, it could have affected operation of the system. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because the likelihood that the system would be operated in the condition that caused the abnormally high vibrations was low, nondestructive examinations revealed no piping degradation, and because no vibrations were observed on the turbine driven auxiliary feedwater train. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200200881, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001007(pdf)
Significance:      Nov 26, 2001 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 6 of 21 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected.
Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C.
The instructions failed to include guidance on how much oil to add to pump bearings following maintenance.
Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution.
This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                          Page 7 of 21 cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Jul 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action for diesel generator overspeed trip switch.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred because the corrective action taken by the licensee regarding the emergency diesel Generator B overspeed trip switch was inadequate. On June 21, 2001, the screws that held the overspeed trip switch intact were found to be loose. The emergency diesel generator had to be removed from service for repair. Repair consisted of tightening the screws that held the switch in place. No other repair action was taken nor was a root cause analysis conducted. On April 9, 2002, the same screws on the same switch were loose and found to be damaged. This also required the emergency diesel generator to be removed from service for repair. Procedure APA-ZZ-00500, "Corrective Action Program," Revision 31, required that a thorough root cause analysis be performed for this level deficiency. The corrective actions taken in response to the first failure, including the failure to perform a root cause analysis, were not adequate to prevent the second failure. This problem identification and resolution finding was more than minor because failure of the overspeed trip switch could have made the diesel generator inoperable. This finding affected the mitigating system cornerstone. The finding was found to be of very low safety significance using the significance determination process because the emergency diesel generator was not determined to be inoperable and the other emergency diesel generator was available. Because this finding was of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Numbers 200103939 and 200202342, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy (Section 40A2.1).
Inspection Report# : 2002002(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone.
This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                          Page 8 of 21 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power.
This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness.
Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 9 of 21 mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
Inspection Report# : 2001003(pdf)
Significance:        Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White) using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2001009(pdf)
Inspection Report# : 2002002(pdf)
Significance:        Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event. Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                      Page 10 of 21 Significance:      Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:      Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:      Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                      Page 11 of 21 Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire.
Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program. Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                      Page 12 of 21 Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:        May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 13 of 21 Significance:        Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Barrier Integrity Significance:        Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                          Page 14 of 21 Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 15 of 21 Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field."
Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance:        Jun 07, 2001 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 16 of 21 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:      Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 17 of 21 methods, and failed to ensure good communications between radiation protection personnel and other work groups.
Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.)
The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:      Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups.
Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:      Sep 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 18 of 21 methods, and failed to ensure good communications between radiation protection personnel and other work groups.
Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems.
The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2001007(pdf)
Inspection Report# : 2000017(pdf)
Significance:      Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded. The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Public Radiation Safety Significance:      Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                      Page 19 of 21 the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7
- Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 - Radioactive Material UN2982."
Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit.
However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 20 of 21 Miscellaneous Significance: N/A Feb 27, 2002 Identified By: NRC Item Type: FIN Finding Deficiencies with implementation of corrective action and operability evaluation programs.
The team determined that several opportunities were missed to promptly identify and correct a risk significant condition adverse to quality involving the degraded condition of the condensate storage tank diaphragm seal. Quality assurance personnel were not actively involved in providing oversight of the event review team and root cause investigation processes. The event review team process did not ensure that statements were obtained from all personnel involved in the event. The corrective action program did not include guidance or expectations on the assignment of appropriate resources to review the highest classification of significant conditions adverse to quality. Minimal resources were initially assigned to the root cause investigation and may have contributed to the delay in identifying the degraded diaphragm seal. Based on interviews with the licensee's staff and a review of the corrective action program procedure, the team determined that licensed operators were only notified of equipment deficiencies if the individual discovering the condition believed there was an immediate impact on nuclear, plant, or personnel safety. Consequently, the potential existed for operability decisions to be made by non-licensed personnel. The operability evaluation program did not implement the guidance provided in NRC Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions."
Inspection Report# : 2002007(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective. However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Callaway                                                                        Page 21 of 21 Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65 (a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/03/2003
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 1 of 20 Callaway Initiating Events Significance:      Jan 12, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadvertent reactor protection system actuation.
During a trip actuating device operational test surveillance, maintenance personnel failed to complete a step in the procedure, resulting in the inadvertent tripping of a reactor trip breaker. This was a violation of Technical Specification 5.4.1. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. Equipment designed to mitigate the consequences of a reactor trip was available and the reactor trip bypass breaker had been closed prior to the inadvertent opening of the reactor trip breaker.
Inspection Report# : 2001002(pdf)
Significance:      Nov 25, 2000 Identified By: Self Disclosing Item Type: FIN Finding Maintenance performed an offsite access circuit without a procedure.
On October 18, 2000, the licensee overhauled a 345 kV switchyard breaker without using a procedure. This breaker was part of the licensee's offsite access circuit. During the overhaul a small fire occurred in the breaker control cabinet.
A significant contributor to the fire was that there was no formal procedure for performing overhaul on switchyard breakers. This finding was determined to have very low safety significance because the lack of procedural guidance for performing maintenance on offsite access circuits did not result in any identified loss of safety or safety support system function and the required offsite sources remained available.
Inspection Report# : 2000015(pdf)
Mitigating Systems Significance: N/A Aug 23, 2002 Identified By: NRC Item Type: FIN Finding Assessment of corrective actions for inoperable auxiliary feedwater pump.
The NRC performed this supplemental inspection to assess the licensee's corrective actions associated with the inoperability of a motor-driven auxiliary feedwater pump. This performance issue was previously characterized as having low to moderate risk significance in NRC Inspection Report 50-483/02-07. During this inspection, the NRC concluded that the licensee had effectively identified and implemented corrective actions for the root and contributing causes for the inoperability of the auxiliary feedwater pump. Based on effective implementation of the corrective actions, it appeared that the inoperability of the pump as a result of foam being entrained in the suction of the pump, was adequately addressed. The effectiveness of the overall corrective action program changes documented in NRC Inspection Report 50-483/02-09, and the licensee's letter to NRC, dated May 8, 2002, will be reviewed during the Problem Identification and Resolution inspection, currently scheduled for December 2002. The performance issue associated with the White inspection finding will remain open until that review is completed.
Inspection Report# : 2002009(pdf)
 
3Q/2002 Inspection Findings - Callaway                                                                          Page 2 of 20 Significance:      Jul 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action for diesel generator overspeed trip switch.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred because the corrective action taken by the licensee regarding the emergency diesel Generator B overspeed trip switch was inadequate. On June 21, 2001, the screws that held the overspeed trip switch intact were found to be loose. The emergency diesel generator had to be removed from service for repair. Repair consisted of tightening the screws that held the switch in place. No other repair action was taken nor was a root cause analysis conducted. On April 9, 2002, the same screws on the same switch were loose and found to be damaged. This also required the emergency diesel generator to be removed from service for repair. Procedure APA-ZZ-00500, "Corrective Action Program," Revision 31, required that a thorough root cause analysis be performed for this level deficiency. The corrective actions taken in response to the first failure, including the failure to perform a root cause analysis, were not adequate to prevent the second failure. This problem identification and resolution finding was more than minor because failure of the overspeed trip switch could have made the diesel generator inoperable. This finding affected the mitigating system cornerstone. The finding was found to be of very low safety significance using the significance determination process because the emergency diesel generator was not determined to be inoperable and the other emergency diesel generator was available. Because this finding was of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Numbers 200103939 and 200202342, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy (Section 40A2.1).
Inspection Report# : 2002002(pdf)
Significance:      Jun 25, 2002 Identified By: NRC Item Type: NCV NonCited Violation Unsecured fire door.
A noncited violation of Operating License Condition 2.C(5)(c) occurred when the licensee failed to take compensatory action when the 3-hour rated fire doors that separated the two trains of control room air conditioning were unlatched and not closed. This finding is more than minor because if a fire had occurred while the doors were unlatched and not closed, they could not perform their function of preventing a fire from spreading from one fire area to another fire area.
This finding affected the mitigating system cornerstone. This finding was evaluated using Appendix F of the reactor safety significance determination process and determined to be of very low safety significance because the combustible load for the area was low and because the fire detectors on each side of the doors were operable. This finding was entered into the licensee's corrective action system as Callaway Action Request System Number 200204041.
Inspection Report# : 2002002(pdf)
Significance:      May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to control design input for degraded voltage relay calculation.
Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, used to determine the degraded voltage relay dropout setting, referred to superseded calculations for important design inputs, and had not been updated to reflect plant configuration and loading changes. This was contrary to the requirement in Procedure EDP-ZZ-04023 that calculations be revised whenever a new or revised calculation (having an effect on the calculation) is issued. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the degraded voltage relay setpoint remained valid). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CARs 200203080 and 200203057), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy.
Inspection Report# : 2002004(pdf)
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 3 of 20 Significance:      May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate calculation of diesel loading.
Requirements in Procedure EDP-ZZ-04023, "Calculations", Revision 14, were not applied correctly to the diesel generator steady-state loading calculation contained in Callaway Drawing E-21005, "List of Loads Supplied by Emergency Diesel Generator," Revision 25. The drawing functioned as a calculation, but lacked the quality requirements for calculations required by this procedure. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the diesel generators retained adequate margin). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CAR 200203017), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy Inspection Report# : 2002004(pdf)
Significance:      May 24, 2002 Identified By: NRC Item Type: FIN Finding Incomplete and incorrect methods to evaluate degraded voltage conditions.
Two licensee calculations contained incomplete and incorrect methods of evaluating degraded voltage conditions.
Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, did not consider the voltage requirements for non-motor loads in determining the degraded voltage relay setting. In addition, Calculation ZZ-214, "Motor Operated Valve Feeder Cable Voltage Drops," Addenda 1, Revision 2, for determining minimum voltage to motor-operated valves, did not consider the effect of motor starting currents in circuit elements upstream of the motor control centers. This finding, which did not involve a violation of NRC requirements, was of very low safety significance because the calculation errors did not result in an actual loss of safety function (the degraded voltage relay setpoint remained valid).
Inspection Report# : 2002004(pdf)
Significance:      Apr 23, 2002 Identified By: Self Disclosing Item Type: FIN Finding Foreign material in condensate transfer system.
A leather weld rod pouch lodged inside the fill valve to the condensate storage tank could have adversely affected the auxiliary feedwater system if the pouch became dislodged while filling the tank. This finding is more than minor because the lack of foreign material exclusion controls could have resulted in the leather weld rod pouch entering the condensate storage tank and adversely affecting the auxiliary feedwater system. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance using the reactor safety significance determination process because no loss of safety function occurred and only one of three auxiliary feedwater pumps would have been affected. This finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200202678.
Inspection Report# : 2002002(pdf)
Significance:      Mar 13, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify the need for and implement corrective action to address the degraded condition of the Auxiliary Feedwater System Train B During the independent review, the team determined that the licensee failed to promptly identify the need for and
 
3Q/2002 Inspection Findings - Callaway                                                                      Page 4 of 20 implement corrective action to address the flow anomaly condition of the auxiliary feedwater system Train B that existed between February 2000 and March 28, 2001, where the flow through the recirculation valve was below the required flow. The condition had a credible impact on safety since the flow anomaly had only been addressed from the standpoint of pump performance and operability and not system performance and required train function. However, since there was no actual loss of safety function and the system would have delivered the required minimum of 500 gpm to two steam generators when the function was required, the finding was considered to be of very low safety significance. Because of the very low safety significance and because the licensee included the item in their corrective action program by reopening Callaway Action Request 200000669 on March 1, 2002, this violation is being treated as a noncited violation (50-483/0208-01) in accordance with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2002008(pdf)
Significance:      Mar 13, 2002 Identified By: NRC Item Type: NCV NonCited Violation Foreign object renders B Essential Service Water pump inoperable A noncited violation of Technical Specification 3.0.3 occurred five times during the time that the Essential Service Water pump was inoperable, three of which exceeded the one hour requirement for initiating actions identified in Technical Specification 3.0.3. Specifically, on February 14, 2001, at 8:51 a.m., the licensee declared the ESW Pump B inoperable due to a tygon tube drain line becoming entwined around the pump impeller. At the same time, Containment Cooler C was out of service for planned maintenance. This met the conditions for entry into TS 3.0.3. The licensee restored the containment cooler to service at 11:15 a.m., which was 2 hours and 32 minutes after when Technical Specification 3.0.3. should have been entered. Four other instances were identified where TS 3.0.3 should have been entered, two of the four times exceeded the one-hour action requirement identified in the TS. Due to the fact that the licensee was unaware that the ESW pump was inoperable from 2:15 p.m. on February 9 until 8:51a.m. on February 14, 2001, they had not realized that they had entered TS 3.0.3 several times. The finding was more than minor because it had an actual impact on safety in that one of the essential service water pumps was rendered inoperable for a duration greater than the allowed outage time while the plant was in a mode of operation that requires the ESW system to be operable. This finding was found to be of very low safety significance because the other train of Essential Service Water was always operable, and there was no actual emergency requiring the operation of the essential service water system. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200100515, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2002008(pdf)
Significance: N/A Mar 13, 2002 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of the event that occurred between February 9 - 15, 2001, where one train of Essential Service Water had been rendered inoperable for approximately 132 hours. If a loss of offsite power had occurred while a train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their safety function. The finding was previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/01-09. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes pertaining to the inoperable Essential Service Water pump and correctly identified the extent of the conditions for having one train of Essential Service Water inoperable for approximately 132 hours.
The licensee's evaluation identified the primary root causes of the performance issues to be: (1) personnel did not know that they needed to secure the drain hose because corrective action from a previous event did not preclude foreign material from entering the suction bay for the essential service water pump, (2) the drain hose was not adequately secured because there was no procedure for the job, (3) the drain hose was not adequately secured because important information that should have been covered during the pre-job brief was omitted, (4) personnel did not know that they needed to secure the drain hose because safety precautions and warnings were left out of the work package, (5) personnel that saw or were informed of the presence of a funnel without a drain hose did not have a questioning
 
3Q/2002 Inspection Findings - Callaway                                                                          Page 5 of 20 attitude, (6) the control room took over one hour to enter Technical Specification 3.0.3 after declaring "B" Essential Service Water system inoperable because personnel found the procedure difficult to use, and (7) the control room took over one hour to enter Technical Specification 3.0.3 after declaring "B" Essential Service Water system inoperable because training was not repeated enough times so that information could be learned and skills practiced. With regard to the extent of condition, the licensee found that the first five root causes identified extended throughout the plant for both installation of leakage control devices and foreign material exclusion. The licensee specified appropriate corrective actions to address the root causes and had implemented these actions by January, 2002. Because of the licensee's acceptable performance in addressing the inoperability of the "B" Essential Service Water system, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2002008(pdf)
Significance:      Feb 27, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to verify calculational methods.
Calculations for auxiliary feedwater pump net positive suction head did not account for nitrogen saturated water. The failure of calculational methods to verify the adequacy of net positive suction head requirements for the auxiliary feedwater pumps was a violation of 10 CFR Part 50, Appendix B, Criterion III. The failure to account for nitrogen saturated water in the net positive suction head calculation for the AFW pumps was more than minor because there was a credible impact on safety in that the available margin of net positive suction head was reduced by 11 feet. Using Phase 1 of the Significance Determination Process, the issue was determined to be of very low safety significance because adequate available net positive suction head remained after accounting for dissolved nitrogen. Therefore, the auxiliary feedwater pump would have remained available during an actual plant event. The finding was entered in the licensee's corrective action program as Callaway Action Report System Item CARS 200200485.
Inspection Report# : 2002007(pdf)
Significance:      Feb 27, 2002 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a significant condition adverse to quality.
Between January 1992 and January 31, 2002, several opportunities were missed to promptly identify and correct a significant condition adverse to quality involving foreign material in the auxiliary feedwater system and condensate storage tank. The failure to promptly identify the degraded condition resulted in the failure of an auxiliary feedwater pump on December 3, 2001. In addition, between January 25 and 29, 2002, the identification of a significant condition adverse to quality involving the as-found condition of the degraded diaphragm seal was not reported to the appropriate levels of management. The multiple examples of missed opportunities to identify a significant condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and also represented a significant human performance cross cutting issue involving the timely recognition of degraded conditions. The finding had greater than minor significance because there was a credible impact on plant safety. Specifically, auxiliary feedwater Pump A failed to run when started by operations personnel during a plant shutdown. Had a plant event occurred, the potential existed for foam from the degraded condensate storage tank diaphragm to fail one or more auxiliary feedwater pumps. The failure of an auxiliary feedwater pump would have adversely affected the decay heat removal critical safety function. A Significance Determination Process Phase 3 analysis preliminarily determined that the issue had low to moderate safety significance (White). This finding was entered in the licensee's corrective action program as Callaway Action Request System Item CARS 200107423.
Inspection Report# : 2002007(pdf)
Significance:      Feb 08, 2002 Identified By: Self Disclosing
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 6 of 20 Item Type: NCV NonCited Violation Inadequate corrective action to address auxiliary feedwater system vibration.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that abnormally high vibration on both motor driven trains of the auxiliary feedwater system was corrected. During the past 12 years, the licensee had identified this condition five times. The licensee did not determine the actual cause of auxiliary feedwater piping vibration and consequently did not take appropriate corrective action. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had a credible impact on safety in that, if this vibration had occurred when auxiliary feedwater was needed, it could have affected operation of the system. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because the likelihood that the system would be operated in the condition that caused the abnormally high vibrations was low, nondestructive examinations revealed no piping degradation, and because no vibrations were observed on the turbine driven auxiliary feedwater train. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200200881, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001007(pdf)
Significance:        Nov 26, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform corrective action.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when a previously identified condition, associated with auxiliary feedwater minimum discharge pressure and recirculation flow, had not been corrected.
Specifically on November 26, 2001, the licensee recognized that, in April 1997 and September 1998, they had identified that the motor-driven auxiliary feedwater pumps had the potential to degrade to a point where they would still be operable in accordance with Technical Specifications, but would not be able to provide the minimum design flow rate to the steam generators. The finding was more than minor because it had an actual impact on safety in that one of the auxiliary feedwater pumps could degrade to a point where it would be operable but unable to perform its design function. This finding was found to be only of very low safety significance because there was no actual degradation of the motor-driven auxiliary feedwater pumps and the turbine-driven auxiliary feedwater pump was available. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107295, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Nov 19, 2001 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to perform adequate maintenance on component cooling water Pump C A noncited violation of Technical Specification 5.4.1 occurred when inadequate maintenance instructions resulted in maintenance personnel not adding enough lubricating oil to the driving bearing of component cooling water Pump C.
The instructions failed to include guidance on how much oil to add to pump bearings following maintenance.
Insufficient lubricating oil caused the pump bearing to fail. This finding is more than minor because it had a credible impact on safety in that, if the other component cooling water pump that supplied the train had failed, the train would not have been available to perform its safety function. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because no other risk significance equipment was rendered inoperable due to the inadequate maintenance instructions and the safety function was still maintained. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200107296, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
 
3Q/2002 Inspection Findings - Callaway                                                                          Page 7 of 20 Significance:        Oct 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take action to ensure emergency core cooling system flood doors were properly controlled.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure flood doors leading into the emergency core cooling system pump rooms were properly controlled. On October 7, 2001, the inspectors identified that the flood door leading to emergency core cooling system Train A equipment was open and unmonitored. With the door open a continuous flood watch was required. In June 2001, the inspectors identified that the flood door leading to emergency core cooling system Train B equipment was open and unmonitored. In response to the June 2001 incident, the licensee did not take corrective action to prevent the doors from being unmonitored while open. The corrective actions for this incident had been closed with no immediate corrective action taken. This finding included crosscutting aspects in the area of problem identification and resolution.
This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was open and unmonitored, fire water could affect the operation of emergency core cooling system equipment. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance because of the low likelihood of a fire water pipe break while the door was open and unmonitored and because of the availability of Train B equipment. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request 200106307, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Flood door left open and unmonitored A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, occurred when the licensee failed to provide continuous monitoring of an open flood door that led into the safety injection pump and centrifugal charging pump Train B areas as required by Engineering Procedure EDP-ZZ-04107, "HVAC Pressure Boundary and Watertight Door Control," Revision 11. This finding is more than minor because it had a credible impact on safety in that, if a fire water pipe break had occurred while the flood door was left open and unmonitored, fire water could affect operation of the safety injection pump and centrifugal charging pump Train B. This finding affects the mitigating system cornerstone.
This finding was found to be only of very low safety significance because of the low likelihood of a fire water pipe break while the flood door was open and unmonitored and because of the availability of Train A equipment. Because this finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200104044, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: FIN Finding Inadequate monitoring of feedwater piping degradation The flow accelerated corrosion program failed to detect degradation in multiple portions of feedwater piping inside the containment building and in the turbine building prior to degradation beyond design minimum wall thickness.
Although the main feedwater degradation was identified and addressed by the licensee before failure, the extent of the degradation at the time of discovery and exposure time while in this condition was a safety concern. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an credible impact on safety and additionally could credibly affect the availability/reliability of a mitigating system (auxiliary feedwater). This finding was determined to be of very low safety significance using the reactor safety significance determination process because the degraded piping was determined to be operable. This issue is in the licensee's corrective action program as Callaway Action Request System Number 200102270.
 
3Q/2002 Inspection Findings - Callaway                                                                          Page 8 of 20 Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action to address turbine driven auxiliary feedwater pump inoperability A noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that the turbine-driven auxiliary feedwater pump's steam trap and adjacent piping were not insulated. Insulating the steam trap and adjacent piping adversely affected the steam trap and caused the pump to become inoperable on June 12, 2001, when condensate level rose to the alarm setpoint while the steam line drain bypass level valve was out of service for maintenance. In August 1994, and on March 19, 2001, an insulated steam trap and/or adjacent piping also caused the turbine-driven auxiliary feedwater pump to become inoperable; however, the licensee failed to take corrective action following these two events to prevent the pump from becoming inoperable on June 12. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had an actual impact on safety in that the turbine-driven auxiliary feedwater pump was rendered inoperable. The event was of very low safety significance because the pump was out of service for less than 4 hours and both motor-driven auxiliary feedwater pumps were available. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200103722, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and manage risk when essential service water was removed from service A noncited violation (EA-01-173) of 10 CFR 50.65(a)(4) occurred when the licensee failed to adequately assess the risk when essential service water Train A was removed from service. Had the risk been adequately assessed, the licensee would have identified that the plant was actually in a higher risk category. The higher risk category required the development of contingency plans to manage the additional risk while essential service water Train A was out of service. This finding is more than minor and had a credible impact on safety because, with essential service water out of service, a diesel generator would not be available to perform its function in the event of a loss of all offsite power.
This placed the plant in a higher risk category and the risk was not adequately assessed or managed. This finding affects the mitigating system cornerstone. This finding was evaluated using Appendix G (Shutdown Operations) of the reactor safety significance determination process and was determined to be of very low safety significance. The minimum equipment required by Appendix G remained available and the other diesel generator was operable. Because this finding is of very low safety significance, and the finding was entered into the licensee corrective action program as Callaway Action Request System Number 200103053, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001003(pdf)
Significance:        Jun 04, 2001 Identified By: NRC Item Type: VIO Violation Essential service water Pump B inoperable for aproximately 132 hours.
On February 9, 2001, a 20-foot section of reinforced tygon hose entered the suction bay of essential service water Pump B, rendering the pump inoperable for approximately 132 hours while the plant operated in Mode 1. Technical Specification 3.7.8.B specified an allowed outage time of 72 hours with the plant in Mode 1, 2, 3, or 4. This is an apparent violation of Technical Specification 3.7.8.B. This finding had greater than minor significance because it had an actual impact on safety, in that a train of essential service water (mitigating system) was inoperable for approximately 132 hours. It has been preliminarily determined to have low to moderate safety significance (White)
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 9 of 20 using the significance determination process worksheet for loss of offsite power. If a loss of offsite power had occurred while the train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their intended functions to mitigate the consequences of the loss of offsite power event. This violation was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 01-0515. The final significance determination for a White finding and a notice of violation were issued for EA-01-130 on July 23, 2001 (ML012050133).
Inspection Report# : 2002002(pdf)
Inspection Report# : 2001009(pdf)
Significance:      Mar 16, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to take Technical Specifications actions for inoperable diesel generators.
The licensee repeatedly failed to enter Technical Specification 3.8.1, Action B.1, while performing Technical Specifications Surveillance Requirement 3.8.1.16. Performance of Technical Specifications Surveillance Requirement 3.8.1.16 involved removal of synchronizing check relays for calibration, which rendered the emergency diesel generators incapable of being synchronized with offsite power sources as required by Technical Specifications Surveillance Requirement 3.8.1.16. The failure to enter Technical Specification 3.8.1, Action B.1, which involved verifying correct breaker alignment and indicated power availability for each required offsite circuit, was first identified by the licensee on August 8, 2000. On December 13, 2000, the licensee identified that this surveillance had been performed six times since August 2000 without performing the required actions. These subsequent events were a result of ineffective corrective action to prevent recurrence and failure to complete a timely root cause analysis for the August 2000 event. This violation of Criterion XVI of 10 CFR Part 50, Appendix B, is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy and was entered into the licensee's corrective action program as Callaway Action Request 00-3135. This noncited violation was characterized as having very low safety significance through the use of the significance determination process. This was because that although the capability to synchronize the emergency diesel generators with offsite power was defeated by removal of the synchronization check relays, they would have properly started and assumed safety-related electrical loads during a loss-of-offsite power event. Also, the licensee determined that none of the times for which the emergency diesel generators were inoperable exceeded the completion time of 1 hour allowed by Technical Specification 3.8.1, Action B.1.
Inspection Report# : 2001004(pdf)
Significance:      Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Ineffective chemistry controls.
The licensee's chemical treatment to plant water systems was ineffective in that it did not control the growth the Asiatic clams in the service water and essential service water systems. As a result, essential service water flow to several safety-related heat exchangers was degraded and flow to the motor-driven auxiliary feedwater Pump A room cooler was reduced below its operability limit. This caused the pump to become inoperable. The failure to establish an adequate chemical treatment program to prevent fouling of heat exchanger surfaces was a violation of Technical Specification 5.4.1. This noncited violation was determined to have very low safety significance because no other safety-related components, other than motor-driven auxiliary feedwater Pump A, was rendered inoperable due to ineffective chemistry controls. The other auxiliary feedwater pumps remained operable.
Inspection Report# : 2000015(pdf)
Significance:      Nov 25, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Motor driven auxiliary feedwater Pump A inoperable due to reduced essential service water flow.
Motor-driven auxiliary feedwater Pump A became inoperable and exceeded its Technical Specification allowed outage
 
3Q/2002 Inspection Findings - Callaway                                                                      Page 10 of 20 time when essential service water flow to the pump room cooler fell below its operability requirement. Flow was reduced to the room cooler due to an Asiatic clam infestation in the essential service system. This was a violation of Technical Specification 3.7.5. This noncited violation was determine to have very low safety significance because, even though Asiatic clams caused the pump to become inoperable, the 100 percent motor-driven auxiliary feedwater Train B and the 200 percent turbine-driven auxiliary feedwater train remained operable. As a result, there was only a small increase in plant risk with the motor-driven auxiliary feedwater Pump A inoperable.
Inspection Report# : 2000015(pdf)
Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of making a change to the fire protection program, without prior Commission approval, that adversely affected the ability to achieve and maintain safe shutdown.
In Fire Area A-27 (reactor trip switchgear room) the team found that redundant equipment required for safe shutdown of the plant following a fire was not separated in accordance with Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1, in that the 20 feet of horizontal space between redundant trains of safe shutdown equipment contained intervening combustibles. Subsequent to this finding, the licensee identified similar conditions in Fire Areas A-1A (west corridor of the 1974 foot elevation of the auxiliary building), and Fire Area A-18 (north electrical penetration room in the auxiliary building). The team also found that in 1989, and 1996, the licensee performed engineering evaluations to justify installed configurations in several fire areas, including Fire Areas A-1A, A-18, and A-27, which did not meet the separation criteria of Section C.5.b of Branch Technical Position Chemical Engineering Branch 9.5-1. In performing these evaluations, however, the licensee failed to consider, as intervening combustibles or fire hazards, non-safety-related cables and other equipment located in the 20 foot separation areas between redundant trains of equipment necessary to achieve and maintain safe shutdown conditions. Therefore, the licensee did not identify the safe shutdown equipment which could be vulnerable to fire damage and the operator actions to restore that equipment to service. The failure to identify and evaluate these additional operator actions were considered by the team to have an adverse affect on the licensee's ability to achieve and maintain safe shutdown in the event of a fire.
Therefore, the team concluded that without prior approval of the Commission, the licensee made changes to their approved fire protection program that adversely affected their ability to achieve and maintain safe shutdown in the event of a fire in Fire Areas A-1A, A-18, and A-27. This is a violation of Operating License Condition 2.C(5)(d), with three examples, and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-2070 and posted compensatory measures in accordance with the provisions of their fire protection program. Each example of this violation was evaluated using the significance determination process, which indicated that, for each of the fire areas involved, the violation had very low safety significance, because the ignition frequencies were relatively low, fire detection and suppression systems were not degraded, and operator actions were available to ensure a safe shutdown path for a fire in each of the fire areas.
Inspection Report# : 2000013(pdf)
Significance:        Aug 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation Noncited violation involving the failure to assure that the design basis was correctly translated into drawings and procedures, and that the adequacy of design was verified or checked-closes URI 0009.
During a previous inspection, NRC inspectors identified an unresolved item involving a potential violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The potential violation concerned the licensee's failure to consider auxiliary feedwater system flow demand on the essential service water system flow balance between 1984 and 1998. The licensee stated that they had not included the auxiliary feedwater flow demand on the essential service water flow balance because they had incorrectly credited the nonsafety-related condensate storage tank as the required water supply for the auxiliary feedwater pumps. The licensee performed a past operability review and determined that the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998. This issue was determined to be a violation of Criterion III of
 
3Q/2002 Inspection Findings - Callaway                                                                      Page 11 of 20 Appendix B to 10 CFR Part 50. This violation is being treated as noncited violation consistent with Section VI.A of the NRC Enforcement Policy. The inspectors determined that the issue had very low safety significance because the essential service water pumps had been capable of supplying adequate flow to the auxiliary feedwater pumps and all other safety-related loads between 1984 and 1998.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain in effect a 3-hour rated fire barrier between redundant trains of equipment necessary to achieve and maintain safe shutdown.
The inspectors identified that a 3-hour rated fire door between the Train A and Train B safety-related ac switchgear rooms was ajar. This failure to properly maintain in effect all provisions of their NRC-approved fire protection program is a violation of Operating License Condition 2.C(5)(c). This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. The licensee entered this finding into their corrective action program as Suggestion-Occurrence-Solution 00-1927. This finding was of very low safety significance, because the door was ajar for less than 3 hours, the ignition frequency was relatively low, and the fire detection and suppression systems were minimally affected.
Inspection Report# : 2000013(pdf)
Significance:        May 26, 2000 Identified By: NRC Item Type: FIN Finding Essential service water system vibration issues were not recognized by licensee personnel in a timely fashion.
During review and closure of Unresolved Item 50-483/0003-01 (essential service water reliability issues), the team noted that licensee personnel had documented several component failures in the essential service water system which were attributable to cyclic stress caused by excessive vibration. These components started failing after implementation of modifications (a May 1992 modification which increased the size of Orifices EFFO0005 and EFFO0006 located in the essential service water return to the ultimate heat sink, and the October 1996 and February 1997 changeout of two system Butterfly Valves EFV0090 and EFV0058). The licensee had not considered either additional vibration or cumulative effects caused by modifications to essential service water, which had experienced high vibration levels since initial plant startup. The team noted that, until May 1999, the licensee had not implemented any significant initiatives to address these issues. At that time, comprehensive corrective actions were finalized, some of which have been implemented. The team concluded after review of the plans, that the licensee is now effectively managing essential service water system vibration and that the reliability of the system should no longer be challenged by vibration. The licensee determined, and the team agreed, that the essential service water system had remained operable throughout this period. Therefore, the team concluded that the vibration issues had a very low risk significance and did not pose a significant safety concern. This issue was determined to be GREEN after being evaluated in the significance determination process.
Inspection Report# : 2000009(pdf)
Significance:        May 25, 2000 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to properly evaluate a plant modification The licensee failed to recognize that a plant modification, which capped two of the four floor drains in Rooms 1206 and 1207 (below the auxiliary feedwater pump rooms), resulted in the facility being outside the design and licensing basis for internal flooding with respect to the consequences of a postulated break in the nonseismic condensate storage tank piping. The team considered this to be a violation of Criterion III of Appendix B to 10 CFR Part 50, which requires assurance that the design basis is correctly translated into drawings and procedures, and that the adequacy of design is verified or checked. This violation is being treated as a Non-Cited Violation (50-483/0009-01), consistent with Section
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 12 of 20 VI.A of the NRC Enforcement Policy. The condition resulting in the violation is in the licensee's corrective action system as Suggestion Occurrence Solution 00-1214 initiated May 25, 2000. This issue was evaluated to have very low risk significance for the safety-related instruments or electrical connections in these rooms because flooding would be limited to approximately 6 inches, which is below the instrumentation installation height. Other equipment in the rooms subject to flooding at this elevation would not be required for safe shutdown.
Inspection Report# : 2000009(pdf)
Significance:      May 20, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures for testing of the turbine driven auxiliary feedwater pump.
The licensee did not comply with the initial condition of a surveillance test procedure requiring that both diesel generators be operable prior to testing the turbine- driven auxiliary feedwater pump. This violation of Technical Specification 6.8.1 is being treated as a noncited violation in accordance with Section VI.A.1 of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 99-3305. The actual risk significance of this issue was very low (Green) because the other diesel generator and its associated 100 percent capacity motor-driven auxiliary feedwater pump were operable and the turbine-driven auxiliary feedwater pump tested satisfactorily.
Inspection Report# : 2000010(pdf)
Significance:      Apr 27, 2000 Identified By: NRC Item Type: FIN Finding Inoperable diesel generator not factored into risk assessment.
The inspectors identified that the plant was in a more risk significant condition than that which was calculated by the risk monitor (quantitative risk assessment) when a diesel generator was made inoperable during maintenance. This placed the plant in the second highest of three risk conditions. The licensee's initial risk assessment did not assume that the diesel generator would be inoperable during maintenance and calculated plant risk as being in the lowest risk condition. Although a qualitative risk assessment performed by operations personnel allowed the diesel generator to be removed from service, it did not indicate that the plant was in a more risk significant configuration and no formal contingency actions were developed. Additionally, the inspectors learned that the licensee's configuration risk monitor program had not defined any contingency actions in response to calculated risk conditions. Failure to account for the diesel generator inoperability in the quantitative risk assessment resulted in the plant being in a more risk-significant condition than most of the plant staff realized. This condition could potentially result in undesirable risk configurations of mitigating systems under certain emergent work situations. However, in this case, other risk-significant equipment was not concurrently removed from service and the error did not result in actual plant risk impact. Therefore, the significance determination process found this issue to be of very low risk significance.
Inspection Report# : 2000010(pdf)
Barrier Integrity Significance:      Jan 10, 2001 Identified By: Self Disclosing Item Type: FIN Finding Unidentified reactor coolant system leakage in excess of Technical Specification limits.
Although operations personnel had prior indication of a valve alignment problem in the boron thermal regeneration system, they were slow to correctly identify the source of the valve alignment problem. As a result, several valves in the boron thermal regeneration system were overpressurized, resulting in reactor coolant system leakage of
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 13 of 20 approximately 2 gpm. This finding was of very low safety significance because once operations personnel identified the valve that was out of alignment they quickly isolated the leak and limited reactor coolant system leakage to approximately 50 gallons.
Inspection Report# : 2001002(pdf)
Significance:        Jun 02, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to comply with the technical specification required action for an inoperable containment penetration An error in a modification package that addressed fire-induced hot short concerns resulted in an outer containment isolation valve (component cooling water return from reactor coolant pump thermal barrier heat exchanger) being inoperable for almost two months. The valve would not have automatically closed on a Phase B (high containment pressure) containment isolation signal. During the time the outer containment isolation valve was inoperable, the inner containment isolation valve for the same penetration was inoperable for 90 minutes. Technical Specification 3.6.3.B required that with both containment isolation valves inoperable that the penetration be isolated within 1 hour. The licensee failed to isolate the penetration as required by Technical Specification 3.6.3.B. This violation of Technical Specification 3.6.3.B is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0314. The actual safety significance of the issue was determined to be very low (Green) because the inner containment isolation valve was inoperable for only 90 minutes. The outer valve could have been remotely closed by a reactor operator from the main control board and the inner valve was not subject to common cause failure because the hot shorts modification had not been performed on it.
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Jul 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise an emergency action level after errors in its bases were identified Inspectors determined that an emergency action level had not been corrected 22 months after licensee staff identified errors in its bases. In March 1998, the licensee determined that there were errors in the calculation of effluent monitor indicators used in determining site area and general emergency classifications. This issue was tracked as Unresolved Item 50-483/00004-02. Subsequently, it was determined to be a violation of 10 CFR 50.54(q) in that the licensee failed to revise an emergency action level associated with plant instrumentation to its most accurate known value to ensure that corresponding protective action recommendations were appropriate for the indicated conditions. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-0108. This issue was of very low safety significance because it did not represent a failure to meet risk significant planning standard 10 CFR 50.47(b)(4) regarding emergency action levels.
Inspection Report# : 2000011(pdf)
Occupational Radiation Safety Significance:        Aug 10, 2001 Identified By: NRC
 
3Q/2002 Inspection Findings - Callaway                                                                          Page 14 of 20 Item Type: NCV NonCited Violation Failure to post a high radiation area.
10 CFR 20.1902(b) requires that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words "Caution High Radiation Area." On May 27, 2001, the licensee identified that a high radiation area located outside in the radwaste yard was not posted. This event is described in the licensee's corrective action program, reference Corrective Action Report 2001-03509. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2001005(pdf)
Significance:        Aug 10, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey On August 9, 2001, the inspector determined that radiation levels on top of the Nukem solid collection system vessel increased from 60 to 180 millirem per hour after the vessel was drained due to a leak. The failure to perform a radiological survey of the vessel after it had been drained, to identify the increased dose rates, is a violation of 10 CFR 20.1501. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Corrective Action Report 2001-04974. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The issue was more than minor because the failure to perform a radiological survey has a credible impact on safety and has the potential for unplanned or unintended dose.
Inspection Report# : 2001005(pdf)
Significance: N/A Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to use NIOSH certified harness straps and belts on all self contained breathing apparatus 10 CFR 20.1703(a) states, in part, that the licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH). From late 1992 to August 2000, self contained breathing apparatus (SCBA) harness straps and belts were used, which were not NIOSH certified for the type of SCBA in use at Callaway, as described in the licensee's corrective action program (Callaway Action Request System Number 200001969). The significance of this violation was determined to be more than minor, because there was a credible impact on a worker's radiation safety and did not affect the cornerstone. There were extenuating circumstances, because the violation was determined to be more than minor.
Inspection Report# : 2001003(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review or evaluate the use of a nonconforming dose rate instrument On April 18, 2001, the inspector identified a survey instrument (RO-2A, SN 2365) which was tagged out of service as nonconforming on April 12, 2001. The description of the nonconformance was, "reading 20 mr/hr in a 100 mr/hr field."
Health Physics Departmental Procedure HDP-ZZ-04000, "Health Physics Instrumentation Program," Revision 16, requires, in part, that a review of the instrument use must be performed within one working day when a dose rate instrument is nonconforming. No review or evaluation had been conducted. The licensee's failure to conduct a review or evaluation of the use of the nonconforming dose rate instrument within one working day was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102148. The significance of this violation was determined to be more than minor, because it could
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 15 of 20 be reasonably viewed as a precursor to a significant event and it involved conditions contrary to licensee procedures which impact instrumentation related to measuring worker dose. This violation was processed through the occupational radiation safety significance determination process and determined to be of very low safety significance, because there was no overexposure, no substantial potential for overexposure because the instrument was removed from service, and the ability to assess dose was not compromised because the technician was wearing dosimetry.
Inspection Report# : 2001003(pdf)
Significance:        Jun 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedural guidance when moving temporary shielding The inspectors identified that temporary shielding in the chemical and volume control system letdown valve cubical had been moved without a review by health physics supervision. Moving lead shielding without health physics supervision review is a violation of Procedure HTP-ZZ-01101 and Technical Specification 5.4.1. Moving lead shielding has a credible impact on safety and the occurrence could have involved a worker's unplanned, unintended dose or potential of such a dose which could have been significantly greater if radiation levels were higher. However, since there was no overexposure or substantial potential for an overexposure and the ability to assess dose was not compromised, the finding is considered to be of very low safety significance. Because of the very low safety significance of the item and because the licensee has included this item in its corrective action program (as CARS 200102390), this procedure violation is being treated as a non-cited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2001008(pdf)
Significance: N/A Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of Refueling Outage 10 job doses that were not as low as is reasonably achievable (ALARA). Three findings were previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/00-17. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes of radiation doses that were not ALARA and correctly identified the extent of the conditions that led to the doses. The doses were identified by the licensee during post-job reviews following Refueling Outage 10. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) management's failure to establish expectations for keeping dose ALARA, (2) management's failure to communicate a priority for keeping doses ALARA, (3) a culture that did not support the ALARA concept, and (4) administrative controls that did not assure documented ALARA concerns would receive proper priority, appropriate consideration, and comprehensive resolution. With regard to the extent of condition, the licensee found that only the fourth root cause extended beyond the radiation protection department. The licensee specified appropriate corrective actions to address the root causes and had implemented most actions by the start of Refueling Outage 11. However, many of the corrective actions were not institutionalized to prevent recurrence of the problems during outages following Refueling Outage 11. The licensee acknowledged this potential problem and entered it into the corrective action program. The licensee was working on separate, broader corrective actions for the fourth root cause. In addition, the licensee intends to conduct effectiveness evaluations of the corrective actions to ensure their effectiveness. Because of the licensee's acceptable performance in addressing job doses that were not ALARA, the White findings associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2001008(pdf)
Significance:        Sep 05, 2000 Identified By: NRC
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 16 of 20 Item Type: VIO Violation Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups.
Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The first of three white fingings associated with the violation of 10 CFR 20.1101(b) involved scaffolding activities which accrued actual doses greater than 25 person-rems.
The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2001007(pdf)
Inspection Report# : 2000017(pdf)
Significance:      Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction methods, and failed to ensure good communications between radiation protection personnel and other work groups.
Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The second of three white fingings associated with the violation of 10 CFR 20.1101(b) involved steam generator eddy current/robotic plugging/stabilizing/electrosleeving activities accrued actual doses greater than 25 person-rems. The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:      Sep 05, 2000 Identified By: NRC Item Type: FIN Finding Failure to Maintain Radiation Doses As-Low-As-Reasonably-Achievable Because of poor planning and preparation, as well as other causes, six jobs that accrued more than 5 person-rems each during Refueling Outage 10 exceeded their projected job doses by more than 50 percent. The licensee scheduled outage activities to reduce the outage duration rather than to reduce dose, failed to properly train workers in dose reduction
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 17 of 20 methods, and failed to ensure good communications between radiation protection personnel and other work groups.
Because of these performance problems and the licensee's history of high collective radiation doses, the NRC identified the issue as a violation of 10 CFR 20.1101(b), which requires that the licensee use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable. Using the Occupational Radiation Safety Significance Determination Process, the NRC determined that the violation was composed of three parts, each of low to moderate risk significance (white). Of the six jobs that exceeded their dose projections by more than 50 percent, two jobs accrued actual doses greater than 25 person-rems. Thus, because the licensee's 3-year rolling average, collective dose exceeded 135 person-rems (but did not exceed 340 person-rems) each was a white finding. In addition, since there were more than two other jobs that accrued more than 5 person-rems (but less than 25 person-rems), these constituted an additional white finding, for a total of three white findings. [The third of three white fingings associated with the violation of 10 CFR 20.1101(b) involved four jobs, each of which accrued actual doses greater than 5 person-rems (steam generator manway covers and inserts removal and installation; health physics support for primary and secondary steam generator activities; foreign object search and retrieval; and reactor coolant pump seal removal and replacement.)
The final significance determination letter and the associated Notice of Violation (EA-00-208) were issued on January 9, 2001.]
Inspection Report# : 2000017(pdf)
Significance:      Aug 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to barricade a high radiation area On May 17, 2000, the licensee identified that a Caution High Radiation Area boundary was moved on the 2000 foot elevation of the radwaste building, and the area was not barricaded for 5 days. The licensee's procedures define a Caution High Radiation Area as an area with dose rates greater than 100 millirems per hour but less than or equal to 1000 millirems per hour at 30 centimeters from a radiation source. Technical Specification 5.7.1.a states, in part, that each entryway to a high radiation area with dose rates not exceeding 1 rem per hour shall be barricaded. The failure to barricade the above area was a violation of Technical Specification 5.7.1.a. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-1139. This issue was determined to have very low safety significance because there was no overexposure or substantial potential for an overexposure to occur.
Inspection Report# : 2000012(pdf)
Public Radiation Safety Significance:      Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform shipping cask leak test requirement prior to shipment.
10 CFR 71.12(c)(2) requires that a licensee who delivers to a carrier for transport licensed material in a package for which a Certificate of Compliance has been issued by the NRC shall comply with the terms and conditions of the Certificate of Compliance as applicable. On December 10, 1999 (Shipment 99-0075) and again on April 25, 2000 (Shipment 00-0022), dewatered bead resin was shipped to the Barnwell Waste Management Facility for disposal using Package USA/9208/B( ) [NuPac Cask Model No 10-142]. In each case, the leak test required by Section 9.b of the Certificate of Compliance was not performed. These events are described in the licensee's corrective action program, reference Callaway Action Requests 2001-166 and 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded and there was no breach of package during transit.
However, it involved a Certificate of Compliance finding resulting in a shipping cask maintenance/use performance deficiency.
 
3Q/2002 Inspection Findings - Callaway                                                                      Page 18 of 20 Inspection Report# : 2001006(pdf)
Significance:        Nov 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide the correct proper shipping name and shipment identification number.
10 CFR 71.5(a) requires that each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on the public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR Parts 170 through 189 appropriate to the mode of transportation. 49 CFR 172.202(a)(1) and (a)(3) require that the shipping description of a hazardous material on the shipping papers must include the proper shipping name prescribed for the material in Column 2 of 49 CFR 172.101, Hazardous Materials Table, and the identification number prescribed for the material as shown in Column 4 of 49 CFR 172.101, Hazardous Materials Table, respectively. On December 10, 1999, the proper shipping name for Shipment 99-0075 was incorrectly determined to be "Radioactive Material, LSA, n.o.s., 7
- Radioactive Material UN2912" instead of "Radioactive Material, n.o.s., 7 - Radioactive Material UN2982."
Therefore, the shipment's hazardous material identification number was also incorrectly assigned as UN2912 instead of UN2982. This event is described in the licensee's corrective action program, reference Callaway Action Request 2001-168. This finding is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Public Radiation Safety Significance Determination Process because radiation limits were not exceeded, and there was no breach of package during transit, certificate of compliance problem, low level burial access problem, or failure to make notifications or provide emergency information.
Inspection Report# : 2001006(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately survey items released from the radiologically controlled area The inspector found that the licensee had not evaluated the ability of its personnel contamination monitors, portable frisking instruments, and tool monitors to identify all radionuclides that might be present on items released from its control. Without this evaluation, the licensee could not ensure that release surveys were adequately performed. The licensee's failure to adequately survey items released from the radiologically controlled area was a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200102126. The significance of this violation was determined to be more than minor, because it could reasonably be viewed as a precursor to a significant event and it involved an occurrence in the radioactive material control program. This violation was processed through the public radiation safety significance determination process and determined to be of very low safety significance, because it did not result in public dose greater than 0.005 rem, and there were no more than five related events Inspection Report# : 2001003(pdf)
Physical Protection Miscellaneous Significance: N/A Feb 27, 2002 Identified By: NRC Item Type: FIN Finding
 
3Q/2002 Inspection Findings - Callaway                                                                        Page 19 of 20 Deficiencies with implementation of corrective action and operability evaluation programs.
The team determined that several opportunities were missed to promptly identify and correct a risk significant condition adverse to quality involving the degraded condition of the condensate storage tank diaphragm seal. Quality assurance personnel were not actively involved in providing oversight of the event review team and root cause investigation processes. The event review team process did not ensure that statements were obtained from all personnel involved in the event. The corrective action program did not include guidance or expectations on the assignment of appropriate resources to review the highest classification of significant conditions adverse to quality. Minimal resources were initially assigned to the root cause investigation and may have contributed to the delay in identifying the degraded diaphragm seal. Based on interviews with the licensee's staff and a review of the corrective action program procedure, the team determined that licensed operators were only notified of equipment deficiencies if the individual discovering the condition believed there was an immediate impact on nuclear, plant, or personnel safety. Consequently, the potential existed for operability decisions to be made by non-licensed personnel. The operability evaluation program did not implement the guidance provided in NRC Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions."
Inspection Report# : 2002007(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Significance: N/A Mar 16, 2001 Identified By: NRC Item Type: FIN Finding Licensee's problem identification and resolution program was effective.
The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments were effective in identifying problems. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. Corrective actions, when specified, were generally implemented in a timely manner. With a few exceptions identified by the licensee, corrective actions to prevent recurrence of conditions adverse to quality were effective. However, one example of untimely and ineffective corrective action, involving testing of emergency diesel generator relays, is discussed as a noncited violation.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Oct 03, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to report the inadvertent start of the diesel generator within the required 4 hours.
On October 3, 2000, while reviewing the procedural guidance for locally starting the diesel generator, a nonlicensed operator started the diesel generator by inadvertently breaking the glass cover for the emergency start button on the local control panel. Operations personnel failed to report the start of the diesel generator as a manual actuation of an engineered safety feature within the 4-hour time requirement. Quality assurance personnel subsequently identified that this condition was reportable. Failing to report the manual actuation of the diesel generator within the required 4 hours was a violation of 10 CFR 50.72(b)(2)(ii). This violation is being treated as a noncited violation consistent with Section
 
3Q/2002 Inspection Findings - Callaway                                                                      Page 20 of 20 VI.A.1 of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Suggestion-Occurrence-Solution Report 00-2450.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Jun 02, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to monitor the performance of a condenser air radiation gas detector Certain cognizant licensee personnel were not aware that a condenser air radiation gas detector was within the scope of the maintenance rule. The detector was identified in the emergency operating procedure to provide an indication of a steam generator tube rupture. Since licensee personnel were not aware the detector was within the scope of the maintenance rule, functional failure determinations had not been performed on detector failures. Without functional failure determinations, the licensee could not demonstrate that the detector was being effectively controlled through preventive maintenance, as required by the maintenance rule. This was a Severity Level IV violation of 10 CFR 50.65 (a)(1) and (2). This violation (EA-00-174) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's correction action program as Suggestion-Occurrence-Solution Report 00-1548. The licensee could still manually sample steam generator blowdown or use other indications of a steam generator tube rupture.
Inspection Report# : 2000011(pdf)
Last modified : December 02, 2002
 
4Q/2002 Inspection Findings - Callaway                                                                                                  Page 1 of 5 Callaway Initiating Events Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate control of over temperature-delta temperature delta flux penalty circuit amplifier gain resulted in a reactor trip.
A noncited violation of 10 CFR Part 50, Appendix B, Criteria III, Design Control, occurred when the licensee failed to maintain control of the over temperature-delta temperature delta flux penalty circuit amplifier gain. The finding was greater than minor because the condition resulted in a transient initiator and contributed to an unplanned reactor trip, an initiating event. This finding was evaluated using Appendix A of the reactor safety significance determination process and determined to be of very low safety significance because the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident, did not contribute to both the likelihood of a reactor trip and the unavailability of mitigation equipment, and did not increase the likelihood of a fire or flood. This finding is in the licensee's corrective action system as Callaway Action Request System Number 200208352.
Inspection Report# : 2002006(pdf)
Mitigating Systems Significance: N/A Aug 23, 2002 Identified By: NRC Item Type: FIN Finding Assessment of corrective actions for inoperable auxiliary feedwater pump.
The NRC performed this supplemental inspection to assess the licensee's corrective actions associated with the inoperability of a motor-driven auxiliary feedwater pump. This performance issue was previously characterized as having low to moderate risk significance in NRC Inspection Report 50-483/02-07. During this inspection, the NRC concluded that the licensee had effectively identified and implemented corrective actions for the root and contributing causes for the inoperability of the auxiliary feedwater pump. Based on effective implementation of the corrective actions, it appeared that the inoperability of the pump as a result of foam being entrained in the suction of the pump, was adequately addressed. The effectiveness of the overall corrective action program changes documented in NRC Inspection Report 50-483/02-09, and the licensee's letter to NRC, dated May 8, 2002, will be reviewed during the Problem Identification and Resolution inspection, currently scheduled for December 2002. The performance issue associated with the White inspection finding will remain open until that review is completed.
Inspection Report# : 2002009(pdf)
Significance:        Jul 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action for diesel generator overspeed trip switch.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred because the corrective action taken by the licensee regarding the emergency diesel Generator B overspeed trip switch was inadequate. On June 21, 2001, the screws that held the overspeed trip switch intact were found to be loose. The emergency diesel generator had to be removed from service for repair. Repair consisted of tightening the screws that held the switch in place. No other repair action was taken nor was a root cause analysis conducted. On April 9, 2002, the same screws on the same switch were loose and found to be damaged. This also required the emergency diesel generator to be removed from service for repair.
Procedure APA-ZZ-00500, "Corrective Action Program," Revision 31, required that a thorough root cause analysis be performed for this level deficiency. The corrective actions taken in response to the first failure, including the failure to perform a root cause analysis, were not adequate to prevent the second failure. This problem identification and resolution finding was more than minor because failure of the overspeed trip switch could have made the diesel generator inoperable. This finding affected the mitigating system cornerstone. The finding was found to be of very low safety significance using the significance determination process because the emergency diesel generator was not determined to be inoperable and the other emergency diesel generator was available. Because this finding was of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Numbers 200103939 and 200202342, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy (Section 40A2.1).
Inspection Report# : 2002002(pdf)
Significance:        Jun 25, 2002
 
4Q/2002 Inspection Findings - Callaway                                                                                                Page 2 of 5 Identified By: NRC Item Type: NCV NonCited Violation Unsecured fire door.
A noncited violation of Operating License Condition 2.C(5)(c) occurred when the licensee failed to take compensatory action when the 3-hour rated fire doors that separated the two trains of control room air conditioning were unlatched and not closed. This finding is more than minor because if a fire had occurred while the doors were unlatched and not closed, they could not perform their function of preventing a fire from spreading from one fire area to another fire area. This finding affected the mitigating system cornerstone. This finding was evaluated using Appendix F of the reactor safety significance determination process and determined to be of very low safety significance because the combustible load for the area was low and because the fire detectors on each side of the doors were operable. This finding was entered into the licensee's corrective action system as Callaway Action Request System Number 200204041.
Inspection Report# : 2002002(pdf)
Significance:        May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to control design input for degraded voltage relay calculation.
Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, used to determine the degraded voltage relay dropout setting, referred to superseded calculations for important design inputs, and had not been updated to reflect plant configuration and loading changes. This was contrary to the requirement in Procedure EDP-ZZ-04023 that calculations be revised whenever a new or revised calculation (having an effect on the calculation) is issued. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the degraded voltage relay setpoint remained valid). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CARs 200203080 and 200203057), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy.
Inspection Report# : 2002004(pdf)
Significance:        May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate calculation of diesel loading.
Requirements in Procedure EDP-ZZ-04023, "Calculations", Revision 14, were not applied correctly to the diesel generator steady-state loading calculation contained in Callaway Drawing E-21005, "List of Loads Supplied by Emergency Diesel Generator," Revision 25. The drawing functioned as a calculation, but lacked the quality requirements for calculations required by this procedure. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the diesel generators retained adequate margin).
Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CAR 200203017), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy Inspection Report# : 2002004(pdf)
Significance:        May 24, 2002 Identified By: NRC Item Type: FIN Finding Incomplete and incorrect methods to evaluate degraded voltage conditions.
Two licensee calculations contained incomplete and incorrect methods of evaluating degraded voltage conditions. Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, did not consider the voltage requirements for non-motor loads in determining the degraded voltage relay setting. In addition, Calculation ZZ-214, "Motor Operated Valve Feeder Cable Voltage Drops," Addenda 1, Revision 2, for determining minimum voltage to motor-operated valves, did not consider the effect of motor starting currents in circuit elements upstream of the motor control centers. This finding, which did not involve a violation of NRC requirements, was of very low safety significance because the calculation errors did not result in an actual loss of safety function (the degraded voltage relay setpoint remained valid).
Inspection Report# : 2002004(pdf)
Significance:        Apr 23, 2002 Identified By: Self Disclosing Item Type: FIN Finding Foreign material in condensate transfer system.
A leather weld rod pouch lodged inside the fill valve to the condensate storage tank could have adversely affected the auxiliary feedwater system if the pouch became dislodged while filling the tank. This finding is more than minor because the lack of foreign material exclusion controls could have resulted in the leather weld rod pouch entering the condensate storage tank and adversely affecting the auxiliary feedwater
 
4Q/2002 Inspection Findings - Callaway                                                                                                  Page 3 of 5 system. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance using the reactor safety significance determination process because no loss of safety function occurred and only one of three auxiliary feedwater pumps would have been affected. This finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200202678.
Inspection Report# : 2002002(pdf)
Significance: N/A Mar 13, 2002 Identified By: NRC Item Type: FIN Finding Supplemental inspection results This supplemental inspection was performed by the NRC to assess the licensee's evaluation of the event that occurred between February 9 - 15, 2001, where one train of Essential Service Water had been rendered inoperable for approximately 132 hours. If a loss of offsite power had occurred while a train of essential service water was inoperable, the Train B safety systems supported by essential service water, including an emergency diesel generator, would not have been available to perform their safety function. The finding was previously characterized as having low to moderate safety significance (White) in NRC Inspection Report 50-483/01-09. During this supplemental inspection performed in accordance with Inspection Procedure 95002, the inspectors determined that the licensee performed a thorough evaluation of the causes pertaining to the inoperable Essential Service Water pump and correctly identified the extent of the conditions for having one train of Essential Service Water inoperable for approximately 132 hours. The licensee's evaluation identified the primary root causes of the performance issues to be: (1) personnel did not know that they needed to secure the drain hose because corrective action from a previous event did not preclude foreign material from entering the suction bay for the essential service water pump, (2) the drain hose was not adequately secured because there was no procedure for the job, (3) the drain hose was not adequately secured because important information that should have been covered during the pre-job brief was omitted, (4) personnel did not know that they needed to secure the drain hose because safety precautions and warnings were left out of the work package, (5) personnel that saw or were informed of the presence of a funnel without a drain hose did not have a questioning attitude, (6) the control room took over one hour to enter Technical Specification 3.0.3 after declaring "B" Essential Service Water system inoperable because personnel found the procedure difficult to use, and (7) the control room took over one hour to enter Technical Specification 3.0.3 after declaring "B" Essential Service Water system inoperable because training was not repeated enough times so that information could be learned and skills practiced. With regard to the extent of condition, the licensee found that the first five root causes identified extended throughout the plant for both installation of leakage control devices and foreign material exclusion. The licensee specified appropriate corrective actions to address the root causes and had implemented these actions by January, 2002. Because of the licensee's acceptable performance in addressing the inoperability of the "B" Essential Service Water system, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters, in accordance with the guidance in IMC 0305, "Operating Reactor Assessment Program." Implementation of the licensee's corrective actions will be reviewed further during a future inspection.
Inspection Report# : 2002008(pdf)
Significance:        Mar 13, 2002 Identified By: NRC Item Type: NCV NonCited Violation Foreign object renders B Essential Service Water pump inoperable A noncited violation of Technical Specification 3.0.3 occurred five times during the time that the Essential Service Water pump was inoperable, three of which exceeded the one hour requirement for initiating actions identified in Technical Specification 3.0.3. Specifically, on February 14, 2001, at 8:51 a.m., the licensee declared the ESW Pump B inoperable due to a tygon tube drain line becoming entwined around the pump impeller. At the same time, Containment Cooler C was out of service for planned maintenance. This met the conditions for entry into TS 3.0.3. The licensee restored the containment cooler to service at 11:15 a.m., which was 2 hours and 32 minutes after when Technical Specification 3.0.3. should have been entered. Four other instances were identified where TS 3.0.3 should have been entered, two of the four times exceeded the one-hour action requirement identified in the TS. Due to the fact that the licensee was unaware that the ESW pump was inoperable from 2:15 p.m. on February 9 until 8:51a.m. on February 14, 2001, they had not realized that they had entered TS 3.0.3 several times. The finding was more than minor because it had an actual impact on safety in that one of the essential service water pumps was rendered inoperable for a duration greater than the allowed outage time while the plant was in a mode of operation that requires the ESW system to be operable. This finding was found to be of very low safety significance because the other train of Essential Service Water was always operable, and there was no actual emergency requiring the operation of the essential service water system. Because the finding is of very low safety significance and the finding was entered into the licensee's corrective action program as Callaway Action Request 200100515, the associated finding is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2002008(pdf)
Significance:        Mar 13, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly identify the need for and implement corrective action to address the degraded condition of the Auxiliary Feedwater System Train B During the independent review, the team determined that the licensee failed to promptly identify the need for and implement corrective action to address the flow anomaly condition of the auxiliary feedwater system Train B that existed between February 2000 and March 28, 2001, where the flow through the recirculation valve was below the required flow. The condition had a credible impact on safety since the flow
 
4Q/2002 Inspection Findings - Callaway                                                                                              Page 4 of 5 anomaly had only been addressed from the standpoint of pump performance and operability and not system performance and required train function. However, since there was no actual loss of safety function and the system would have delivered the required minimum of 500 gpm to two steam generators when the function was required, the finding was considered to be of very low safety significance. Because of the very low safety significance and because the licensee included the item in their corrective action program by reopening Callaway Action Request 200000669 on March 1, 2002, this violation is being treated as a noncited violation (50-483/0208-01) in accordance with Section VI.A.1 of the NRC Enforcement Policy.
Inspection Report# : 2002008(pdf)
Significance:        Feb 27, 2002 Identified By: NRC Item Type: VIO Violation Failure to promptly identify and correct a significant condition adverse to quality.
Between January 1992 and January 31, 2002, several opportunities were missed to promptly identify and correct a significant condition adverse to quality involving foreign material in the auxiliary feedwater system and condensate storage tank. The failure to promptly identify the degraded condition resulted in the failure of an auxiliary feedwater pump on December 3, 2001. In addition, between January 25 and 29, 2002, the identification of a significant condition adverse to quality involving the as-found condition of the degraded diaphragm seal was not reported to the appropriate levels of management. The multiple examples of missed opportunities to identify a significant condition adverse to quality was a violation of 10 CFR Part 50, Appendix B, Criterion XVI and also represented a significant human performance cross cutting issue involving the timely recognition of degraded conditions. The finding had greater than minor significance because there was a credible impact on plant safety. Specifically, auxiliary feedwater Pump A failed to run when started by operations personnel during a plant shutdown. Had a plant event occurred, the potential existed for foam from the degraded condensate storage tank diaphragm to fail one or more auxiliary feedwater pumps. The failure of an auxiliary feedwater pump would have adversely affected the decay heat removal critical safety function. A Significance Determination Process Phase 3 analysis preliminarily determined that the issue had low to moderate safety significance (White).
This finding was entered in the licensee's corrective action program as Callaway Action Request System Item CARS 200107423.
Inspection Report# : 2002007(pdf)
Significance:        Feb 27, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to verify calculational methods.
Calculations for auxiliary feedwater pump net positive suction head did not account for nitrogen saturated water. The failure of calculational methods to verify the adequacy of net positive suction head requirements for the auxiliary feedwater pumps was a violation of 10 CFR Part 50, Appendix B, Criterion III. The failure to account for nitrogen saturated water in the net positive suction head calculation for the AFW pumps was more than minor because there was a credible impact on safety in that the available margin of net positive suction head was reduced by 11 feet. Using Phase 1 of the Significance Determination Process, the issue was determined to be of very low safety significance because adequate available net positive suction head remained after accounting for dissolved nitrogen. Therefore, the auxiliary feedwater pump would have remained available during an actual plant event. The finding was entered in the licensee's corrective action program as Callaway Action Report System Item CARS 200200485.
Inspection Report# : 2002007(pdf)
Significance:        Feb 08, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate corrective action to address auxiliary feedwater system vibration.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred when the licensee failed to take corrective action to ensure that abnormally high vibration on both motor driven trains of the auxiliary feedwater system was corrected. During the past 12 years, the licensee had identified this condition five times. The licensee did not determine the actual cause of auxiliary feedwater piping vibration and consequently did not take appropriate corrective action. This finding included crosscutting aspects in the area of problem identification and resolution. The finding was more than minor because it had a credible impact on safety in that, if this vibration had occurred when auxiliary feedwater was needed, it could have affected operation of the system. This finding affects the mitigating system cornerstone. This finding was found to be only of very low safety significance because the likelihood that the system would be operated in the condition that caused the abnormally high vibrations was low, nondestructive examinations revealed no piping degradation, and because no vibrations were observed on the turbine driven auxiliary feedwater train. Because the finding is of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200200881, the associated violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy Inspection Report# : 2001007(pdf)
 
4Q/2002 Inspection Findings - Callaway                                                                                                Page 5 of 5 Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Significance: N/A Feb 27, 2002 Identified By: NRC Item Type: FIN Finding Deficiencies with implementation of corrective action and operability evaluation programs.
The team determined that several opportunities were missed to promptly identify and correct a risk significant condition adverse to quality involving the degraded condition of the condensate storage tank diaphragm seal. Quality assurance personnel were not actively involved in providing oversight of the event review team and root cause investigation processes. The event review team process did not ensure that statements were obtained from all personnel involved in the event. The corrective action program did not include guidance or expectations on the assignment of appropriate resources to review the highest classification of significant conditions adverse to quality. Minimal resources were initially assigned to the root cause investigation and may have contributed to the delay in identifying the degraded diaphragm seal. Based on interviews with the licensee's staff and a review of the corrective action program procedure, the team determined that licensed operators were only notified of equipment deficiencies if the individual discovering the condition believed there was an immediate impact on nuclear, plant, or personnel safety. Consequently, the potential existed for operability decisions to be made by non-licensed personnel. The operability evaluation program did not implement the guidance provided in NRC Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions."
Inspection Report# : 2002007(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : March 25, 2003
 
1Q/2003 Inspection Findings - Callaway                                                                            Page 1 of 6 Callaway 1Q/2003 Plant Inspection Findings Initiating Events Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate control of over temperature-delta temperature delta flux penalty circuit amplifier gain resulted in a reactor trip.
A noncited violation of 10 CFR Part 50, Appendix B, Criteria III, Design Control, occurred when the licensee failed to maintain control of the over temperature-delta temperature delta flux penalty circuit amplifier gain. The finding was greater than minor because the condition resulted in a transient initiator and contributed to an unplanned reactor trip, an initiating event. This finding was evaluated using Appendix A of the reactor safety significance determination process and determined to be of very low safety significance because the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident, did not contribute to both the likelihood of a reactor trip and the unavailability of mitigation equipment, and did not increase the likelihood of a fire or flood. This finding is in the licensee's corrective action system as Callaway Action Request System Number 200208352.
Inspection Report# : 2002006(pdf)
Mitigating Systems Significance:        Mar 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure of the turbine-driven auxiliary feed pump due to incorrectly manufactured and installed part.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to correctly manufacture and install a valve stem on the turbine-driven auxiliary feedwater turbine. Appropriate quantitative and qualitative measures were not utilized to assure that the valve stem was manufactured to the correct dimensions, as required by Appendix B, prior to installation. This finding had actual safety significance because the condition resulted in the failure of the turbine-driven auxiliary feedwater pump to respond to a valid demand signal.
The finding was more than minor because it was associated with the mitigating system equipment performance cornerstone attribute and adversely affected the availability/reliability cornerstone objective. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003003(pdf)
Significance: N/A Aug 23, 2002 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Callaway                                                                            Page 2 of 6 Item Type: FIN Finding Assessment of corrective actions for inoperable auxiliary feedwater pump.
The NRC performed this supplemental inspection to assess the licensee's corrective actions associated with the inoperability of a motor-driven auxiliary feedwater pump. This performance issue was previously characterized as having low to moderate risk significance in NRC Inspection Report 50-483/02-07. During this inspection, the NRC concluded that the licensee had effectively identified and implemented corrective actions for the root and contributing causes for the inoperability of the auxiliary feedwater pump. Based on effective implementation of the corrective actions, it appeared that the inoperability of the pump as a result of foam being entrained in the suction of the pump, was adequately addressed. The effectiveness of the overall corrective action program changes documented in NRC Inspection Report 50-483/02-09, and the licensee's letter to NRC, dated May 8, 2002, will be reviewed during the Problem Identification and Resolution inspection, currently scheduled for December 2002. The performance issue associated with the White inspection finding will remain open until that review is completed.
Inspection Report# : 2002009(pdf)
Significance:      Jul 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action for diesel generator overspeed trip switch.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred because the corrective action taken by the licensee regarding the emergency diesel Generator B overspeed trip switch was inadequate. On June 21, 2001, the screws that held the overspeed trip switch intact were found to be loose. The emergency diesel generator had to be removed from service for repair. Repair consisted of tightening the screws that held the switch in place. No other repair action was taken nor was a root cause analysis conducted. On April 9, 2002, the same screws on the same switch were loose and found to be damaged. This also required the emergency diesel generator to be removed from service for repair. Procedure APA-ZZ-00500, "Corrective Action Program," Revision 31, required that a thorough root cause analysis be performed for this level deficiency. The corrective actions taken in response to the first failure, including the failure to perform a root cause analysis, were not adequate to prevent the second failure. This problem identification and resolution finding was more than minor because failure of the overspeed trip switch could have made the diesel generator inoperable. This finding affected the mitigating system cornerstone. The finding was found to be of very low safety significance using the significance determination process because the emergency diesel generator was not determined to be inoperable and the other emergency diesel generator was available. Because this finding was of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Numbers 200103939 and 200202342, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy (Section 40A2.1).
Inspection Report# : 2002002(pdf)
Significance:      Jun 25, 2002 Identified By: NRC Item Type: NCV NonCited Violation Unsecured fire door.
A noncited violation of Operating License Condition 2.C(5)(c) occurred when the licensee failed to take compensatory action when the 3-hour rated fire doors that separated the two trains of control room air conditioning were unlatched and not closed. This finding is more than minor because if a fire had occurred while the doors were unlatched and not closed, they could not perform their function of preventing a fire from spreading from one fire area to another fire area.
This finding affected the mitigating system cornerstone. This finding was evaluated using Appendix F of the reactor safety significance determination process and determined to be of very low safety significance because the combustible load for the area was low and because the fire detectors on each side of the doors were operable. This finding was entered into the licensee's corrective action system as Callaway Action Request System Number 200204041.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Callaway                                                                          Page 3 of 6 Inspection Report# : 2002002(pdf)
Significance:      May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate calculation of diesel loading.
Requirements in Procedure EDP-ZZ-04023, "Calculations", Revision 14, were not applied correctly to the diesel generator steady-state loading calculation contained in Callaway Drawing E-21005, "List of Loads Supplied by Emergency Diesel Generator," Revision 25. The drawing functioned as a calculation, but lacked the quality requirements for calculations required by this procedure. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the diesel generators retained adequate margin). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CAR 200203017), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy Inspection Report# : 2002004(pdf)
Significance:      May 24, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to control design input for degraded voltage relay calculation.
Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, used to determine the degraded voltage relay dropout setting, referred to superseded calculations for important design inputs, and had not been updated to reflect plant configuration and loading changes. This was contrary to the requirement in Procedure EDP-ZZ-04023 that calculations be revised whenever a new or revised calculation (having an effect on the calculation) is issued. The failure to follow procedural requirements was identified as a violation of Criterion V to 10 CFR Part 50, Appendix B, "Instructions, Procedures, and Drawings." This finding was of very low safety significance since there was no actual loss of safety function (the degraded voltage relay setpoint remained valid). Because of the low safety significance and the licensee's action to place the issue in their corrective action program (CARs 200203080 and 200203057), this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy.
Inspection Report# : 2002004(pdf)
Significance:      May 24, 2002 Identified By: NRC Item Type: FIN Finding Incomplete and incorrect methods to evaluate degraded voltage conditions.
Two licensee calculations contained incomplete and incorrect methods of evaluating degraded voltage conditions.
Calculation E-B-21, "LSELS Degraded Voltage Setpoint Calculation," Revision 0, did not consider the voltage requirements for non-motor loads in determining the degraded voltage relay setting. In addition, Calculation ZZ-214, "Motor Operated Valve Feeder Cable Voltage Drops," Addenda 1, Revision 2, for determining minimum voltage to motor-operated valves, did not consider the effect of motor starting currents in circuit elements upstream of the motor control centers. This finding, which did not involve a violation of NRC requirements, was of very low safety significance because the calculation errors did not result in an actual loss of safety function (the degraded voltage relay setpoint remained valid).
Inspection Report# : 2002004(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Callaway                                                                            Page 4 of 6 Significance:        Apr 23, 2002 Identified By: Self Disclosing Item Type: FIN Finding Foreign material in condensate transfer system.
A leather weld rod pouch lodged inside the fill valve to the condensate storage tank could have adversely affected the auxiliary feedwater system if the pouch became dislodged while filling the tank. This finding is more than minor because the lack of foreign material exclusion controls could have resulted in the leather weld rod pouch entering the condensate storage tank and adversely affecting the auxiliary feedwater system. This finding affects the mitigating system cornerstone. This finding was found to be of very low safety significance using the reactor safety significance determination process because no loss of safety function occurred and only one of three auxiliary feedwater pumps would have been affected. This finding was entered into the licensee's corrective action program as Callaway Action Request System Number 200202678.
Inspection Report# : 2002002(pdf)
Barrier Integrity Significance:        Jan 08, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to isolate an inoperable containment penetration flow path.
A noncited violation of Technical Specification Action 3.6.3, Containment Isolation Valves, occurred when the licensee failed to isolate an inoperable component cooling water containment penetration flow path within the prescribed 4 hours. This finding had actual safety significance because it resulted in one of two automatic containment isolation engineering features to be disabled and would have become a more significant safety condition if left uncorrected. This finding was more than minor because it was associated with barrier performance, the containment isolation reliability cornerstone attribute, and adversely affected the barrier integrity cornerstone objective. This finding was evaluated using Appendix A of the reactor safety significance determination process and determined to be of very low safety significance because the condition did not affect the control room barrier function or represent an actual open containment pathway.
Inspection Report# : 2003003(pdf)
Emergency Preparedness Significance: TBD Mar 21, 2003 Identified By: NRC Item Type: AV Apparent Violation Failure to meet the Alert Notification System design criteria due to programmatic deficiencies resulting in an inaccurate Tone Alert Radio database in apparent violation of 10 CFR 50.47(b)(5).
Between September 1998, and November 2002, due to programmatic inadequacies, a small percentage of residences in the licensee's plume exposure emergency planning zone would not have received an emergency alerting signal in the event of an emergency at the Callaway facility. The failure to establish a means to notify members of the public in the emergency planning zone was a violation of 10 CFR 50.47(b)(5), and also represented an apparent human performance cross cutting issue involving the timely recognition and correction of degraded conditions. The finding had greater than file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - Callaway                                                                          Page 5 of 6 minor significance because the condition resulted in a loss of alert notification capability to about 1.5 percent of the emergency planning zone population, and if left uncorrected the condition would have continued to degrade. Using the Emergency Preparedness Significance Determination Process the finding was preliminarily determined to have low to moderate safety significance (White) because it was a violation of 10 CFR 50.47(b)(5) and represented a risk-significant planning standard degraded function failure. This finding was entered in the licensee's corrective action program as Callaway Action Request System Item CARS 200208007.
Inspection Report# : 2003008(pdf)
Occupational Radiation Safety Significance:        Feb 13, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiological surveys.
Inspectors identified two examples of a violation of 10 CFR 20.1501(a) for failure to perform radiological surveys. The licensee failed to collect airborne samples to evaluate the potential for airborne activity during the removal and reinstallation of contaminated insulation on Valve BB8378A on October 29 and November 15, 2002, respectively. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200300355. The issue was more than minor because the failure to perform a radiological survey has the potential for unplanned or unintended dose which could have been significantly greater as a result of higher levels of airborne activity. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because it did not involve ALARA planning and controls, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not compromise the licensee's ability to assess dose.
Inspection Report# : 2003003(pdf)
Public Radiation Safety Physical Protection Significance: N/A Feb 14, 2003 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations.
Inspection Report# : 2003002(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Callaway                                                                          Page 6 of 6 Miscellaneous Significance: N/A Jan 30, 2003 Identified By: NRC Item Type: FIN Finding Implementation of identification and resolution of problems program Issues associated with a failure to identify and adequately evaluate an operability issue associated with the auxiliary feedwater system and two examples of inadequate corrective actions for conditions adverse to quality provided indications that the licensee had weaknesses in their problem identification and resolution program. The team found the licensee effectively implemented changes to address these problem identification and resolution program weaknesses.
Problems were identified at the proper threshold and entered into the corrective action program. Risk information was effectively used to prioritize the extent of evaluation and to determine the schedule for implementation of corrective actions. Corrective actions, when specified, were typically implemented in a timely manner. During interviews workers indicated no reluctance to place safety issues into the problem identification and resolution program. However, a licensee survey indicated that some employees felt that they had received negative repercussions for raising issues.
Inspection Report# : 2002003(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : May 30, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          07/22/2003
 
2Q/2003 Inspection Findings - Callaway                                                                          Page 1 of 6 Callaway 2Q/2003 Plant Inspection Findings Initiating Events Significance:        Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate control of over temperature-delta temperature delta flux penalty circuit amplifier gain resulted in a reactor trip.
A noncited violation of 10 CFR Part 50, Appendix B, Criteria III, Design Control, occurred when the licensee failed to maintain control of the over temperature-delta temperature delta flux penalty circuit amplifier gain. The finding was greater than minor because the condition resulted in a transient initiator and contributed to an unplanned reactor trip, an initiating event. This finding was evaluated using Appendix A of the reactor safety significance determination process and determined to be of very low safety significance because the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident, did not contribute to both the likelihood of a reactor trip and the unavailability of mitigation equipment, and did not increase the likelihood of a fire or flood. This finding is in the licensee's corrective action system as Callaway Action Request System Number 200208352.
Inspection Report# : 2002006(pdf)
Mitigating Systems Significance:        Jun 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct recurring air voiding condition on containment spray system.
The inspectors concluded that voiding of the containment spray suction header occurred on two occasions during the inspection period. The voiding occurred because the licensee failed to properly fill and vent the suction piping following maintenance. The inspectors concluded this condition was a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, and was a finding of very low safety significance. This finding had actual safety significance because the condition resulted in repeated air voiding of a safety-related pump. This finding was greater than minor because it was similar to Example 2C of Appendix E of Inspection Manual Chapter 0612 (i.e., a repetitive issue involving degradation of a safety-related pump). This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003004(pdf)
Significance:        Jun 21, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - Callaway                                                                            Page 2 of 6 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to correctly install a pressurizer safety valve inlet gasket due to inadequate work instructions.
The inspectors concluded that the pressurizer safety valve seat leakage, and subsequent plant shutdown, was the result of incorrect valve reassembly during the previous refueling outage. The inspectors concluded that this condition was a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, and a finding of very low safety significance. This finding was greater than minor because it was associated with the mitigating system equipment performance cornerstone attributes and it affected the availability/reliability cornerstone objective. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003004(pdf)
Significance:        Jun 06, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement effective corrective actions.
The licensee failed to promptly identify, correct, or preclude recurrence of an industry known potential significant condition adverse to quality associated with failures of Magne-Blast 4160 Volt circuit breakers. The breaker failures were the result of a defective contact block assembly used as control switches in the breaker control circuits. The failure to promptly identify, correct, or preclude recurrence of the deficient condition from affecting multiple safety related components due to failures of Magne-Blast 4160 volt circuit breakers was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because if left uncorrected this condition impacts the reliability and availability of all safety related loads supplied by Magne-Blast 4160 Volt circuit breakers.
This finding was determined to be of very low safety significance since all failures reviewed did not result in loss of a safety function for a single train for greater than its Technical Specification allowed outage time.
Inspection Report# : 2003010(pdf)
Significance:        Mar 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure of the turbine-driven auxiliary feed pump due to incorrectly manufactured and installed part.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to correctly manufacture and install a valve stem on the turbine-driven auxiliary feedwater turbine. Appropriate quantitative and qualitative measures were not utilized to assure that the valve stem was manufactured to the correct dimensions, as required by Appendix B, prior to installation. This finding had actual safety significance because the condition resulted in the failure of the turbine-driven auxiliary feedwater pump to respond to a valid demand signal.
The finding was more than minor because it was associated with the mitigating system equipment performance cornerstone attribute and adversely affected the availability/reliability cornerstone objective. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003003(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Callaway                                                                            Page 3 of 6 Significance: N/A Aug 23, 2002 Identified By: NRC Item Type: FIN Finding Assessment of corrective actions for inoperable auxiliary feedwater pump.
The NRC performed this supplemental inspection to assess the licensee's corrective actions associated with the inoperability of a motor-driven auxiliary feedwater pump. This performance issue was previously characterized as having low to moderate risk significance in NRC Inspection Report 50-483/02-07. During this inspection, the NRC concluded that the licensee had effectively identified and implemented corrective actions for the root and contributing causes for the inoperability of the auxiliary feedwater pump. Based on effective implementation of the corrective actions, it appeared that the inoperability of the pump as a result of foam being entrained in the suction of the pump, was adequately addressed. The effectiveness of the overall corrective action program changes documented in NRC Inspection Report 50-483/02-09, and the licensee's letter to NRC, dated May 8, 2002, will be reviewed during the Problem Identification and Resolution inspection, currently scheduled for December 2002. The performance issue associated with the White inspection finding will remain open until that review is completed.
Inspection Report# : 2002009(pdf)
Significance:      Jul 06, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective action for diesel generator overspeed trip switch.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, occurred because the corrective action taken by the licensee regarding the emergency diesel Generator B overspeed trip switch was inadequate. On June 21, 2001, the screws that held the overspeed trip switch intact were found to be loose. The emergency diesel generator had to be removed from service for repair. Repair consisted of tightening the screws that held the switch in place. No other repair action was taken nor was a root cause analysis conducted. On April 9, 2002, the same screws on the same switch were loose and found to be damaged. This also required the emergency diesel generator to be removed from service for repair. Procedure APA-ZZ-00500, "Corrective Action Program," Revision 31, required that a thorough root cause analysis be performed for this level deficiency. The corrective actions taken in response to the first failure, including the failure to perform a root cause analysis, were not adequate to prevent the second failure. This problem identification and resolution finding was more than minor because failure of the overspeed trip switch could have made the diesel generator inoperable. This finding affected the mitigating system cornerstone. The finding was found to be of very low safety significance using the significance determination process because the emergency diesel generator was not determined to be inoperable and the other emergency diesel generator was available. Because this finding was of very low safety significance, and the finding was entered into the licensee's corrective action program as Callaway Action Request System Numbers 200103939 and 200202342, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy (Section 40A2.1).
Inspection Report# : 2002002(pdf)
Barrier Integrity Significance:      Jan 08, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to isolate an inoperable containment penetration flow path.
A noncited violation of Technical Specification Action 3.6.3, Containment Isolation Valves, occurred when the file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Callaway                                                                            Page 4 of 6 licensee failed to isolate an inoperable component cooling water containment penetration flow path within the prescribed 4 hours. This finding had actual safety significance because it resulted in one of two automatic containment isolation engineering features to be disabled and would have become a more significant safety condition if left uncorrected. This finding was more than minor because it was associated with barrier performance, the containment isolation reliability cornerstone attribute, and adversely affected the barrier integrity cornerstone objective. This finding was evaluated using Appendix A of the reactor safety significance determination process and determined to be of very low safety significance because the condition did not affect the control room barrier function or represent an actual open containment pathway.
Inspection Report# : 2003003(pdf)
Emergency Preparedness Significance:        Mar 21, 2003 Identified By: NRC Item Type: VIO Violation Failure to meet the Alert Notification System design criteria due to programmatic deficiencies resulting in an inaccurate Tone Alert Radio database in apparent violation of 10 CFR 50.47(b)(5).
Between September 1998, and November 2002, due to programmatic inadequacies, a small percentage of residences in the licensee's plume exposure emergency planning zone would not have received an emergency alerting signal in the event of an emergency at the Callaway facility. The failure to establish a means to notify members of the public in the emergency planning zone was a violation of 10 CFR 50.47(b)(5), and also involved cross cutting aspects in the area of problem identification. The finding had greater than minor significance because the condition, if left uncorrected, would have continued to degrade resulting in additional loss of alert notification capability. A Significance Determination Process review determined that the issue had low to moderate safety significance (White). The finding was entered in the licensee's corrective action program as Callaway Action Request System Item CARS 200208007.
The final significance determination (White) and Notice of Violation were transmitted to the licensee in a {{letter dated|date=June 20, 2003|text=letter dated June 20, 2003}}.
Inspection Report# : 2003008(pdf)
Occupational Radiation Safety Significance:        Feb 13, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiological surveys.
Inspectors identified two examples of a violation of 10 CFR 20.1501(a) for failure to perform radiological surveys. The licensee failed to collect airborne samples to evaluate the potential for airborne activity during the removal and reinstallation of contaminated insulation on Valve BB8378A on October 29 and November 15, 2002, respectively. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200300355. The issue was more than minor because the failure to perform a radiological survey has the potential for unplanned or unintended dose which could have been significantly greater as a result of higher levels of airborne activity. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - Callaway                                                                          Page 5 of 6 Determination Process because it did not involve ALARA planning and controls, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not compromise the licensee's ability to assess dose.
Inspection Report# : 2003003(pdf)
Public Radiation Safety Physical Protection Significance: N/A Feb 14, 2003 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations.
Inspection Report# : 2003002(pdf)
Miscellaneous Significance: N/A Jun 06, 2003 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems On the basis of the sample selected for review, the team concluded that in general, problems were adequately identified, evaluated, and corrected. The team identified a number of examples pertaining to the failure to promptly identify and correct conditions adverse to quality. One long-standing issue involving a failure to promptly identify and correct voided conditions affecting both trains of the containment spray system suction piping following abnormal system response during surveillance testing on multiple occasions dating back to 1995 was identified by the team.
Problem identification and resolution issues have affected Callaway historically and corrective actions have been put in place to improve performance. The team noted that engineering products reviewed effectively supported the corrective action process, were technically adequate, and provided sufficient justification to support operability for degraded conditions evaluated.
Inspection Report# : 2003010(pdf)
Significance: N/A Jan 30, 2003 Identified By: NRC Item Type: FIN Finding Implementation of identification and resolution of problems program Issues associated with a failure to identify and adequately evaluate an operability issue associated with the auxiliary feedwater system and two examples of inadequate corrective actions for conditions adverse to quality provided file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        10/08/2003
 
2Q/2003 Inspection Findings - Callaway                                                                          Page 6 of 6 indications that the licensee had weaknesses in their problem identification and resolution program. The team found the licensee effectively implemented changes to address these problem identification and resolution program weaknesses.
Problems were identified at the proper threshold and entered into the corrective action program. Risk information was effectively used to prioritize the extent of evaluation and to determine the schedule for implementation of corrective actions. Corrective actions, when specified, were typically implemented in a timely manner. During interviews workers indicated no reluctance to place safety issues into the problem identification and resolution program. However, a licensee survey indicated that some employees felt that they had received negative repercussions for raising issues.
Inspection Report# : 2002003(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999. In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          10/08/2003
 
3Q/2003 Inspection Findings - Callaway                                                                            Page 1 of 7 Callaway 3Q/2003 Plant Inspection Findings Initiating Events Significance:      Dec 28, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate control of over temperature-delta temperature delta flux penalty circuit amplifier gain resulted in a reactor trip.
A noncited violation of 10 CFR Part 50, Appendix B, Criteria III, Design Control, occurred when the licensee failed to maintain control of the over temperature-delta temperature delta flux penalty circuit amplifier gain.
The finding was greater than minor because the condition resulted in a transient initiator and contributed to an unplanned reactor trip, an initiating event. This finding was evaluated using Appendix A of the reactor safety significance determination process and determined to be of very low safety significance because the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident, did not contribute to both the likelihood of a reactor trip and the unavailability of mitigation equipment, and did not increase the likelihood of a fire or flood. This finding is in the licensee's corrective action system as Callaway Action Request System Number 200208352.
Inspection Report# : 2002006(pdf)
Mitigating Systems Significance:      Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions following an EDG rocker arm lube oil valve mispositioning.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action."
This violation was related to inadequate corrective actions taken following an emergency diesel generator rocker arm lube oil valve mispositioning. The licensee's corrective actions were not adequate to prevent recurrence.
This finding was greater than minor because it could reasonably be viewed as a precursor to a significant event and if left uncorrected, would become a more significant safety concern. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003005(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Callaway                                                                            Page 2 of 7 Significance:      Jun 21, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to correctly install a pressurizer safety valve inlet gasket due to inadequate work instructions.
The inspectors concluded that the pressurizer safety valve seat leakage, and subsequent plant shutdown, was the result of incorrect valve reassembly during the previous refueling outage. The inspectors concluded that this condition was a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, and a finding of very low safety significance.
This finding was greater than minor because it was associated with the mitigating system equipment performance cornerstone attributes and it affected the availability/reliability cornerstone objective. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003004(pdf)
Significance:      Jun 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct recurring air voiding condition on containment spray system.
The inspectors concluded that voiding of the containment spray suction header occurred on two occasions during the inspection period. The voiding occurred because the licensee failed to properly fill and vent the suction piping following maintenance. The inspectors concluded this condition was a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, and was a finding of very low safety significance.
This finding had actual safety significance because the condition resulted in repeated air voiding of a safety-related pump. This finding was greater than minor because it was similar to Example 2C of Appendix E of Inspection Manual Chapter 0612 (i.e., a repetitive issue involving degradation of a safety-related pump). This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003004(pdf)
Significance:      Jun 06, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement effective corrective actions.
The licensee failed to promptly identify, correct, or preclude recurrence of an industry known potential significant condition adverse to quality associated with failures of Magne-Blast 4160 Volt circuit breakers. The breaker failures were the result of a defective contact block assembly used as control switches in the breaker control circuits.
The failure to promptly identify, correct, or preclude recurrence of the deficient condition from affecting multiple safety related components due to failures of Magne-Blast 4160 volt circuit breakers was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Callaway                                                                            Page 3 of 7 Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because if left uncorrected this condition impacts the reliability and availability of all safety related loads supplied by Magne-Blast 4160 Volt circuit breakers. This finding was determined to be of very low safety significance since all failures reviewed did not result in loss of a safety function for a single train for greater than its Technical Specification allowed outage time.
Inspection Report# : 2003010(pdf)
Significance:        Mar 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure of the turbine-driven auxiliary feed pump due to incorrectly manufactured and installed part.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to correctly manufacture and install a valve stem on the turbine-driven auxiliary feedwater turbine. Appropriate quantitative and qualitative measures were not utilized to assure that the valve stem was manufactured to the correct dimensions, as required by Appendix B, prior to installation.
This finding had actual safety significance because the condition resulted in the failure of the turbine-driven auxiliary feedwater pump to respond to a valid demand signal. The finding was more than minor because it was associated with the mitigating system equipment performance cornerstone attribute and adversely affected the availability/reliability cornerstone objective. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003003(pdf)
Barrier Integrity Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure of containment radiation monitors to meet Technical Specifications operability requirements.
The inspectors identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, associated with the licensee's failure to assure that applicable regulatory requirements and the design basis for the containment radiation gas monitors were correctly translated into Calculation GT-13 and, ultimately, the radiation monitor setpoint. This deficiency resulted in the containment gaseous channel becoming incapable of performing the design bases function to detect a one gallon per minute reactor coolant system leak within one hour in accordance with the licensee's commitment to Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems."
This finding was greater than minor because the containment gas channel radiation monitor was not capable of performing the design basis function for an extended period of time. The inoperability of the radiation monitor resulted in potential impact on reactor safety and adversely affected the reactor coolant leakage performance attribute of the barrier integrity reactor safety cornerstone. The finding was only of very low safety significance because other methods of reactor coolant system leak detection were available to the licensee. The unavailability of the gaseous channel leak detection function did not contribute to an increase in core damage sequences when evaluated using the significance determination process Phase 2 worksheets.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Callaway                                                                          Page 4 of 7 Inspection Report# : 2003005(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective actions following an unanalyzed condition.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action."
This violation was related to inadequate corrective actions taken following identification of an unanalyzed condition (control room ventilation envelope door open) which resulted in the postulated postaccident control room dose limits to be exceeded. The licensee's corrective actions failed to prevent recurrence of the condition.
This finding was greater than minor because it was associated with the integrity of the control room envelope. Because this finding involved the degradation of barrier integrity, the finding was evaluated using the significance determination process for at-power situations. The inspectors concluded that the finding was only of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2003005(pdf)
Significance:        Jan 08, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to isolate an inoperable containment penetration flow path.
A noncited violation of Technical Specification Action 3.6.3, Containment Isolation Valves, occurred when the licensee failed to isolate an inoperable component cooling water containment penetration flow path within the prescribed 4 hours.
This finding had actual safety significance because it resulted in one of two automatic containment isolation engineering features to be disabled and would have become a more significant safety condition if left uncorrected. This finding was more than minor because it was associated with barrier performance, the containment isolation reliability cornerstone attribute, and adversely affected the barrier integrity cornerstone objective. This finding was evaluated using Appendix A of the reactor safety significance determination process and determined to be of very low safety significance because the condition did not affect the control room barrier function or represent an actual open containment pathway.
Inspection Report# : 2003003(pdf)
Emergency Preparedness Significance:        Mar 21, 2003 Identified By: NRC Item Type: VIO Violation Failure to meet the Alert Notification System design criteria due to programmatic deficiencies resulting in an inaccurate Tone Alert Radio database in apparent violation of 10 CFR 50.47(b)(5).
Between September 1998, and November 2002, due to programmatic inadequacies, a small percentage of residences in the licensee's plume exposure emergency planning zone would not have received an emergency alerting signal in the event of an emergency at the Callaway facility. The failure to establish a means to notify members of the public in the file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        01/12/2004
 
3Q/2003 Inspection Findings - Callaway                                                                          Page 5 of 7 emergency planning zone was a violation of 10 CFR 50.47(b)(5), and also involved cross cutting aspects in the area of problem identification.
The finding had greater than minor significance because the condition, if left uncorrected, would have continued to degrade resulting in additional loss of alert notification capability. A Significance Determination Process review determined that the issue had low to moderate safety significance (White). The finding was entered in the licensee's corrective action program as Callaway Action Request System Item CARS 200208007. The final significance determination (White) and Notice of Violation were transmitted to the licensee in a {{letter dated|date=June 20, 2003|text=letter dated June 20, 2003}}.
Inspection Report# : 2003012(pdf)
Inspection Report# : 2003008(pdf)
Occupational Radiation Safety Significance:        Feb 13, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiological surveys.
Inspectors identified two examples of a violation of 10 CFR 20.1501(a) for failure to perform radiological surveys. The licensee failed to collect airborne samples to evaluate the potential for airborne activity during the removal and reinstallation of contaminated insulation on Valve BB8378A on October 29 and November 15, 2002, respectively. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200300355.
The issue was more than minor because the failure to perform a radiological survey has the potential for unplanned or unintended dose which could have been significantly greater as a result of higher levels of airborne activity. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because it did not involve ALARA planning and controls, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not compromise the licensee's ability to assess dose.
Inspection Report# : 2003003(pdf)
Public Radiation Safety Significance:        Jul 02, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Dose rates on the external surface of a package in excess of DOT limits.
The licensee failed to maintain contact dose rates to 200 millirems per hour or less on a package transported in an open, exclusive use shipment, in violation of 49 CFR 173.441(b)(1).
This self-revealing, noncited violation was greater than minor because the finding is associated with one of the Public Radiation Safety Cornerstone attributes (transportation packaging) and the finding affects the associated cornerstone objective (to ensure adequate protection of public health and safety from exposure to radioactive materials released into file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - Callaway                                                                          Page 6 of 7 the public domain). The finding was related to an occurrence in the licensee's radioactive material transportation program that was contrary to Department of Transportation regulations and, therefore, was processed through the Public Radiation Safety Significance Determination Process. The finding is of very low safety significance because it involved a radiation dose limit (200 millirems per hour) that was exceeded, but the dose rate (300 millirems per hour) did not exceed the limit by more than two times and it was not accessible to the public.
Inspection Report# : 2003009(pdf)
Physical Protection Significance: N/A Feb 14, 2003 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations.
Inspection Report# : 2003002(pdf)
Miscellaneous Significance: N/A Jun 06, 2003 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems On the basis of the sample selected for review, the team concluded that in general, problems were adequately identified, evaluated, and corrected. The team identified a number of examples pertaining to the failure to promptly identify and correct conditions adverse to quality. One long-standing issue involving a failure to promptly identify and correct voided conditions affecting both trains of the containment spray system suction piping following abnormal system response during surveillance testing on multiple occasions dating back to 1995 was identified by the team.
Problem identification and resolution issues have affected Callaway historically and corrective actions have been put in place to improve performance. The team noted that engineering products reviewed effectively supported the corrective action process, were technically adequate, and provided sufficient justification to support operability for degraded conditions evaluated.
Inspection Report# : 2003010(pdf)
Significance: N/A Jan 30, 2003 Identified By: NRC Item Type: FIN Finding Implementation of identification and resolution of problems program Issues associated with a failure to identify and adequately evaluate an operability issue associated with the auxiliary feedwater system and two examples of inadequate corrective actions for conditions adverse to quality provided indications that the licensee had weaknesses in their problem identification and resolution program. The team found the licensee effectively implemented changes to address these problem identification and resolution program weaknesses.
Problems were identified at the proper threshold and entered into the corrective action program. Risk information was file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        01/12/2004
 
3Q/2003 Inspection Findings - Callaway                                                                          Page 7 of 7 effectively used to prioritize the extent of evaluation and to determine the schedule for implementation of corrective actions. Corrective actions, when specified, were typically implemented in a timely manner. During interviews workers indicated no reluctance to place safety issues into the problem identification and resolution program. However, a licensee survey indicated that some employees felt that they had received negative repercussions for raising issues.
Inspection Report# : 2002003(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999.
In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          01/12/2004
 
4Q/2003 Inspection Findings - Callaway                                                                          Page 1 of 9 Callaway 4Q/2003 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding The failure of a licensed operator to follow a procedure resulted in an unplanned plant transient.
An unplanned plant transient resulted from the failure of an operator to follow a written procedure. The transient occurred after the unexpected loss of all plant service cooling water and all but one of the condenser circulating water pumps. Cooling water was lost after an operator inadvertently opened the feeder breaker supplying power to the pumps.
This finding is greater than minor because the operator error affected the human performance attribute of the initiating events cornerstone. The inspectors determined that the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood. The finding was similar to Example 4.b in MC 0612, Appendix E and was entered into the licensee's corrective action program as Callaway Action Request (CAR) 200308178.
Inspection Report# : 2003006(pdf)
Mitigating Systems Significance:      Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Required Compensatory Measures When CREVIS Operation Rendered ESF Switchgear Room Halon System Inoperable The licensee did not recognize that the halon system protecting both engineered safety feature switchgear rooms was rendered inoperable and, therefore, failed to take the required compensatory action when the control room emergency ventilation and isolation system was in operation. Two ventilation dampers in parallel through the common fire wall between these rooms open when this system starts. The team identified that these dampers do not automatically shut when the halon system actuates. The halon system would not be capable of reaching the required concentration to suppress a fire because halon would be allowed to escape under these conditions. License Condition 2.C.(5)(c) requires that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Standardized Nuclear Unit Power Plant System Final Safety Analysis Report. Updated Final Safety Analysis Report, Table 9.5.1-2, "Halon Systems," requires that when this halon system is inoperable, the licensee shall establish a continuous fire watch with backup fire suppression capability in the affected area. Contrary to this, on numerous occasions throughout the operating life of the plant, the team found that the licensee had failed to post a continuous fire watch whenever the vital switchgear room halon system was rendered inoperable due to testing of the control room ventilation system. This violation of License Condition 2.C.(5)(c) will be treated as a noncited violation, consistent file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Callaway                                                                            Page 2 of 9 with Section VI.A of the Enforcement Policy. This issue was in the licensee's corrective action program under Callaway Action Request 200307189.
This finding was greater than minor because it involved the potential degradation of a fire protection feature protecting the electrical distribution equipment powering both trains of mitigating systems. This finding is of very low safety significance because the fire ignition frequency in the rooms affected is low, the remaining fire detection and suppression capability are unaffected, and sufficient accident mitigation equipment was available.
Inspection Report# : 2003007(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions following an EDG rocker arm lube oil valve mispositioning.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action."
This violation was related to inadequate corrective actions taken following an emergency diesel generator rocker arm lube oil valve mispositioning. The licensee's corrective actions were not adequate to prevent recurrence.
This finding was greater than minor because it could reasonably be viewed as a precursor to a significant event and if left uncorrected, would become a more significant safety concern. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003005(pdf)
Significance:        Jun 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct recurring air voiding condition on containment spray system.
The inspectors concluded that voiding of the containment spray suction header occurred on two occasions during the inspection period. The voiding occurred because the licensee failed to properly fill and vent the suction piping following maintenance. The inspectors concluded this condition was a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, and was a finding of very low safety significance.
This finding had actual safety significance because the condition resulted in repeated air voiding of a safety-related pump. This finding was greater than minor because it was similar to Example 2C of Appendix E of Inspection Manual Chapter 0612 (i.e., a repetitive issue involving degradation of a safety-related pump). This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003004(pdf)
Significance:        Jun 21, 2003 Identified By: Self Disclosing file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - Callaway                                                                            Page 3 of 9 Item Type: NCV NonCited Violation Failure to correctly install a pressurizer safety valve inlet gasket due to inadequate work instructions.
The inspectors concluded that the pressurizer safety valve seat leakage, and subsequent plant shutdown, was the result of incorrect valve reassembly during the previous refueling outage. The inspectors concluded that this condition was a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, and a finding of very low safety significance.
This finding was greater than minor because it was associated with the mitigating system equipment performance cornerstone attributes and it affected the availability/reliability cornerstone objective. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003004(pdf)
Significance:      Jun 06, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement effective corrective actions.
The licensee failed to promptly identify, correct, or preclude recurrence of an industry known potential significant condition adverse to quality associated with failures of Magne-Blast 4160 Volt circuit breakers. The breaker failures were the result of a defective contact block assembly used as control switches in the breaker control circuits.
The failure to promptly identify, correct, or preclude recurrence of the deficient condition from affecting multiple safety related components due to failures of Magne-Blast 4160 volt circuit breakers was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because if left uncorrected this condition impacts the reliability and availability of all safety related loads supplied by Magne-Blast 4160 Volt circuit breakers. This finding was determined to be of very low safety significance since all failures reviewed did not result in loss of a safety function for a single train for greater than its Technical Specification allowed outage time.
Inspection Report# : 2003010(pdf)
Significance:      Mar 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure of the turbine-driven auxiliary feed pump due to incorrectly manufactured and installed part.
A noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to correctly manufacture and install a valve stem on the turbine-driven auxiliary feedwater turbine. Appropriate quantitative and qualitative measures were not utilized to assure that the valve stem was manufactured to the correct dimensions, as required by Appendix B, prior to installation.
This finding had actual safety significance because the condition resulted in the failure of the turbine-driven auxiliary feedwater pump to respond to a valid demand signal. The finding was more than minor because it was associated with the mitigating system equipment performance cornerstone attribute and adversely affected the availability/reliability cornerstone objective. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - Callaway                                                                            Page 4 of 9 potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003003(pdf)
Barrier Integrity Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate incorporation of design information into work instructions lead to the failure of a pressurizer block valve.
The inspectors identified a finding and NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This finding is related to inadequate incorporation of design information into the work instructions for modifications to a pressurizer PORV block valve actuator circuit. The inadequate work instructions resulted in the failure of the valve actuator following return to service.
This finding is greater than minor because the block valve failure affected the reactor coolant system equipment and barrier performance attribute of the barrier integrity cornerstone. The inspectors evaluated the condition with the Phase 2 worksheet because the finding involved the reactor coolant system barrier. This finding is only of very low safety significance because the block valve inoperability did not significantly contribute to an increase in core damage frequency. The licensee placed this issue in their corrective action program as CAR 200306563.
Inspection Report# : 2003006(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate postmaintenance test of a pressurizer power operated relief block valve.
The inspectors identified a finding and noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control."
This finding is related to inadequate testing of the pressurizer power operated relief valve (PORV) block valve following modifications to the actuator circuit. The testing failed to detect that the valve actuator had failed.
This finding is greater than minor because the block valve failure affected the reactor coolant system equipment and barrier performance attribute of the barrier integrity cornerstone. The inspectors evaluated the condition with the Phase 2 worksheet because the finding involved the reactor coolant system barrier. The finding was only of very low safety significance because the block valve failure did not significantly contribute to an increase in core damage frequency.
The licensee placed this issue in their corrective action program as CAR 200306563.
Inspection Report# : 2003006(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective actions following an unanalyzed condition.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action."
This violation was related to inadequate corrective actions taken following identification of an unanalyzed condition (control room ventilation envelope door open) which resulted in the postulated postaccident control room dose limits to file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - Callaway                                                                          Page 5 of 9 be exceeded. The licensee's corrective actions failed to prevent recurrence of the condition.
This finding was greater than minor because it was associated with the integrity of the control room envelope. Because this finding involved the degradation of barrier integrity, the finding was evaluated using the significance determination process for at-power situations. The inspectors concluded that the finding was only of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2003005(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure of containment radiation monitors to meet Technical Specifications operability requirements.
The inspectors identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, associated with the licensee's failure to assure that applicable regulatory requirements and the design basis for the containment radiation gas monitors were correctly translated into Calculation GT-13 and, ultimately, the radiation monitor setpoint. This deficiency resulted in the containment gaseous channel becoming incapable of performing the design bases function to detect a one gallon per minute reactor coolant system leak within one hour in accordance with the licensee's commitment to Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems."
This finding was greater than minor because the containment gas channel radiation monitor was not capable of performing the design basis function for an extended period of time. The inoperability of the radiation monitor resulted in potential impact on reactor safety and adversely affected the reactor coolant leakage performance attribute of the barrier integrity reactor safety cornerstone. The finding was only of very low safety significance because other methods of reactor coolant system leak detection were available to the licensee. The unavailability of the gaseous channel leak detection function did not contribute to an increase in core damage sequences when evaluated using the significance determination process Phase 2 worksheets.
Inspection Report# : 2003005(pdf)
Significance:        Jan 08, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to isolate an inoperable containment penetration flow path.
A noncited violation of Technical Specification Action 3.6.3, Containment Isolation Valves, occurred when the licensee failed to isolate an inoperable component cooling water containment penetration flow path within the prescribed 4 hours.
This finding had actual safety significance because it resulted in one of two automatic containment isolation engineering features to be disabled and would have become a more significant safety condition if left uncorrected. This finding was more than minor because it was associated with barrier performance, the containment isolation reliability cornerstone attribute, and adversely affected the barrier integrity cornerstone objective. This finding was evaluated using Appendix A of the reactor safety significance determination process and determined to be of very low safety significance because the condition did not affect the control room barrier function or represent an actual open containment pathway.
Inspection Report# : 2003003(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        04/22/2004
 
4Q/2003 Inspection Findings - Callaway                                                                        Page 6 of 9 Emergency Preparedness Significance:      Mar 21, 2003 Identified By: NRC Item Type: VIO Violation Failure to meet the Alert Notification System design criteria due to programmatic deficiencies resulting in an inaccurate Tone Alert Radio database in apparent violation of 10 CFR 50.47(b)(5).
Between September 1998, and November 2002, due to programmatic inadequacies, a small percentage of residences in the licensee's plume exposure emergency planning zone would not have received an emergency alerting signal in the event of an emergency at the Callaway facility. The failure to establish a means to notify members of the public in the emergency planning zone was a violation of 10 CFR 50.47(b)(5), and also involved cross cutting aspects in the area of problem identification.
The finding had greater than minor significance because the condition, if left uncorrected, would have continued to degrade resulting in additional loss of alert notification capability. A Significance Determination Process review determined that the issue had low to moderate safety significance (White). The finding was entered in the licensee's corrective action program as Callaway Action Request System Item CARS 200208007. The final significance determination (White) and Notice of Violation were transmitted to the licensee in a {{letter dated|date=June 20, 2003|text=letter dated June 20, 2003}}.
The NRC performed a supplemental inspection to assess the licensee's evaluation associated with the failure to meet requirements of 10 CFR 50.47(b)(5), in that the licensee did not establish a means to notify members of the public in the emergency planning zone of an emergency using tone alert radios in areas lacking effective siren coverage. This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report 50-483/03-08. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee performed a satisfactory evaluation of the White finding. The licensee's evaluation identified the primary root causes of the performance issue to be: (1) situations that were not covered in procedures, (2) inadequate supervision of the Senior Nuclear Clerks, and (3) turn-over processes for the Senior Nuclear Clerks required improvement.
Given the licensee's acceptable performance in addressing the issue, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in Inspection Manual Chapter 0305, "Operating Reactor Assessment Program." The issue was identified in the first quarter of 2003, therefore it will no longer be considered in assessing plant performance after the fourth quarter of 2003.
Inspection Report# : 2003008(pdf)
Inspection Report# : 2003012(pdf)
Occupational Radiation Safety Significance:      Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to barricade a high radiation area.
The inspectors identified a non-cited violation of Technical Specification 5.7.1 because the licensee failed to barricade a high radiation area to prevent inadvertent entry. Specifically, on October 21, 2003, while performing independent file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        04/22/2004
 
4Q/2003 Inspection Findings - Callaway                                                                          Page 7 of 9 radiation measurements, the inspectors identified a high radiation area on the 2031-foot elevation of the radwaste building that was not enclosed by a barricade. Radiation dose rates around a demineralizer sample panel drain tank were as high as 140 millirems per hour at 30 centimeters from the surface penetrated by the radiation. The finding is in the licensee's corrective action program as CAR 200307676.
This finding was greater than minor because inadequate controls of high radiation areas affect the licensee's ability to ensure adequate protection of worker health and safety from exposure to radiation and affected the cornerstone attribute/exposure control. Because the finding involved the potential for workers to receive significant unplanned, unintended dose as a result of conditions contrary to technical specification requirements, the inspector used the Occupational Radiation Safety Significance Determination Process described in Manual Chapter 0609, Appendix C, to analyze the significance of the finding. The inspector determined that a substantial potential for overexposure did not exist; therefore, the finding had very low significance.
Inspection Report# : 2003006(pdf)
Significance:        Feb 13, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiological surveys.
Inspectors identified two examples of a violation of 10 CFR 20.1501(a) for failure to perform radiological surveys. The licensee failed to collect airborne samples to evaluate the potential for airborne activity during the removal and reinstallation of contaminated insulation on Valve BB8378A on October 29 and November 15, 2002, respectively. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Callaway Action Request System Number 200300355.
The issue was more than minor because the failure to perform a radiological survey has the potential for unplanned or unintended dose which could have been significantly greater as a result of higher levels of airborne activity. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety Significance Determination Process because it did not involve ALARA planning and controls, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not compromise the licensee's ability to assess dose.
Inspection Report# : 2003003(pdf)
Public Radiation Safety Significance:        Jul 02, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Dose rates on the external surface of a package in excess of DOT limits.
The licensee failed to maintain contact dose rates to 200 millirems per hour or less on a package transported in an open, exclusive use shipment, in violation of 49 CFR 173.441(b)(1).
This self-revealing, noncited violation was greater than minor because the finding is associated with one of the Public Radiation Safety Cornerstone attributes (transportation packaging) and the finding affects the associated cornerstone objective (to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain). The finding was related to an occurrence in the licensee's radioactive material transportation file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Callaway                                                                          Page 8 of 9 program that was contrary to Department of Transportation regulations and, therefore, was processed through the Public Radiation Safety Significance Determination Process. The finding is of very low safety significance because it involved a radiation dose limit (200 millirems per hour) that was exceeded, but the dose rate (300 millirems per hour) did not exceed the limit by more than two times and it was not accessible to the public.
Inspection Report# : 2003009(pdf)
Physical Protection Significance: N/A Feb 14, 2003 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations.
Inspection Report# : 2003002(pdf)
Miscellaneous Significance: N/A Jun 06, 2003 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems On the basis of the sample selected for review, the team concluded that in general, problems were adequately identified, evaluated, and corrected. The team identified a number of examples pertaining to the failure to promptly identify and correct conditions adverse to quality. One long-standing issue involving a failure to promptly identify and correct voided conditions affecting both trains of the containment spray system suction piping following abnormal system response during surveillance testing on multiple occasions dating back to 1995 was identified by the team.
Problem identification and resolution issues have affected Callaway historically and corrective actions have been put in place to improve performance. The team noted that engineering products reviewed effectively supported the corrective action process, were technically adequate, and provided sufficient justification to support operability for degraded conditions evaluated.
Inspection Report# : 2003010(pdf)
Significance: N/A Jan 30, 2003 Identified By: NRC Item Type: FIN Finding Implementation of identification and resolution of problems program Issues associated with a failure to identify and adequately evaluate an operability issue associated with the auxiliary feedwater system and two examples of inadequate corrective actions for conditions adverse to quality provided indications that the licensee had weaknesses in their problem identification and resolution program. The team found the licensee effectively implemented changes to address these problem identification and resolution program weaknesses.
Problems were identified at the proper threshold and entered into the corrective action program. Risk information was effectively used to prioritize the extent of evaluation and to determine the schedule for implementation of corrective file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                        04/22/2004
 
4Q/2003 Inspection Findings - Callaway                                                                          Page 9 of 9 actions. Corrective actions, when specified, were typically implemented in a timely manner. During interviews workers indicated no reluctance to place safety issues into the problem identification and resolution program. However, a licensee survey indicated that some employees felt that they had received negative repercussions for raising issues.
Inspection Report# : 2002003(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999.
In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CALL\call_pim.html                                                          04/22/2004
 
1Q/2004 Inspection Findings - Callaway                                                                                                Page 1 of 5 Callaway 1Q/2004 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding The failure of a licensed operator to follow a procedure resulted in an unplanned plant transient.
An unplanned plant transient resulted from the failure of an operator to follow a written procedure. The transient occurred after the unexpected loss of all plant service cooling water and all but one of the condenser circulating water pumps. Cooling water was lost after an operator inadvertently opened the feeder breaker supplying power to the pumps.
This finding is greater than minor because the operator error affected the human performance attribute of the initiating events cornerstone. The inspectors determined that the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood. The finding was similar to Example 4.b in MC 0612, Appendix E and was entered into the licensee's corrective action program as Callaway Action Request (CAR) 200308178.
Inspection Report# : 2003006(pdf)
Mitigating Systems Significance:        Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Required Compensatory Measures When CREVIS Operation Rendered ESF Switchgear Room Halon System Inoperable The licensee did not recognize that the halon system protecting both engineered safety feature switchgear rooms was rendered inoperable and, therefore, failed to take the required compensatory action when the control room emergency ventilation and isolation system was in operation.
Two ventilation dampers in parallel through the common fire wall between these rooms open when this system starts. The team identified that these dampers do not automatically shut when the halon system actuates. The halon system would not be capable of reaching the required concentration to suppress a fire because halon would be allowed to escape under these conditions. License Condition 2.C.(5)(c) requires that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Standardized Nuclear Unit Power Plant System Final Safety Analysis Report. Updated Final Safety Analysis Report, Table 9.5.1-2, "Halon Systems," requires that when this halon system is inoperable, the licensee shall establish a continuous fire watch with backup fire suppression capability in the affected area. Contrary to this, on numerous occasions throughout the operating life of the plant, the team found that the licensee had failed to post a continuous fire watch whenever the vital switchgear room halon system was rendered inoperable due to testing of the control room ventilation system. This violation of License Condition 2.C.(5)(c) will be treated as a noncited violation, consistent with Section VI.A of the Enforcement Policy. This issue was in the licensee's corrective action program under Callaway Action Request 200307189.
This finding was greater than minor because it involved the potential degradation of a fire protection feature protecting the electrical distribution equipment powering both trains of mitigating systems. This finding is of very low safety significance because the fire ignition frequency in the rooms affected is low, the remaining fire detection and suppression capability are unaffected, and sufficient accident mitigation equipment was available.
Inspection Report# : 2003007(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions following an EDG rocker arm lube oil valve mispositioning.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation was related to inadequate corrective actions taken following an emergency diesel generator rocker arm lube oil valve mispositioning. The licensee's corrective actions were not adequate to prevent recurrence.
07/14/2004
 
1Q/2004 Inspection Findings - Callaway                                                                                                    Page 2 of 5 This finding was greater than minor because it could reasonably be viewed as a precursor to a significant event and if left uncorrected, would become a more significant safety concern. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003005(pdf)
Significance:        Jun 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct recurring air voiding condition on containment spray system.
The inspectors concluded that voiding of the containment spray suction header occurred on two occasions during the inspection period. The voiding occurred because the licensee failed to properly fill and vent the suction piping following maintenance. The inspectors concluded this condition was a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, and was a finding of very low safety significance.
This finding had actual safety significance because the condition resulted in repeated air voiding of a safety-related pump. This finding was greater than minor because it was similar to Example 2C of Appendix E of Inspection Manual Chapter 0612 (i.e., a repetitive issue involving degradation of a safety-related pump). This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003004(pdf)
Significance:        Jun 21, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to correctly install a pressurizer safety valve inlet gasket due to inadequate work instructions.
The inspectors concluded that the pressurizer safety valve seat leakage, and subsequent plant shutdown, was the result of incorrect valve reassembly during the previous refueling outage. The inspectors concluded that this condition was a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, and a finding of very low safety significance.
This finding was greater than minor because it was associated with the mitigating system equipment performance cornerstone attributes and it affected the availability/reliability cornerstone objective. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003004(pdf)
Significance:        Jun 06, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement effective corrective actions.
The licensee failed to promptly identify, correct, or preclude recurrence of an industry known potential significant condition adverse to quality associated with failures of Magne-Blast 4160 Volt circuit breakers. The breaker failures were the result of a defective contact block assembly used as control switches in the breaker control circuits.
The failure to promptly identify, correct, or preclude recurrence of the deficient condition from affecting multiple safety related components due to failures of Magne-Blast 4160 volt circuit breakers was determined to be a violation of 10 CFR Part 50, Appendix B, Criterion XVI. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This finding is greater than minor because if left uncorrected this condition impacts the reliability and availability of all safety related loads supplied by Magne-Blast 4160 Volt circuit breakers. This finding was determined to be of very low safety significance since all failures reviewed did not result in loss of a safety function for a single train for greater than its Technical Specification allowed outage time.
Inspection Report# : 2003010(pdf)
Barrier Integrity 07/14/2004
 
1Q/2004 Inspection Findings - Callaway                                                                                                Page 3 of 5 Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate postmaintenance test of a pressurizer power operated relief block valve.
The inspectors identified a finding and noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control." This finding is related to inadequate testing of the pressurizer power operated relief valve (PORV) block valve following modifications to the actuator circuit. The testing failed to detect that the valve actuator had failed.
This finding is greater than minor because the block valve failure affected the reactor coolant system equipment and barrier performance attribute of the barrier integrity cornerstone. The inspectors evaluated the condition with the Phase 2 worksheet because the finding involved the reactor coolant system barrier. The finding was only of very low safety significance because the block valve failure did not significantly contribute to an increase in core damage frequency. The licensee placed this issue in their corrective action program as CAR 200306563.
Inspection Report# : 2003006(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate incorporation of design information into work instructions lead to the failure of a pressurizer block valve.
The inspectors identified a finding and NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This finding is related to inadequate incorporation of design information into the work instructions for modifications to a pressurizer PORV block valve actuator circuit.
The inadequate work instructions resulted in the failure of the valve actuator following return to service.
This finding is greater than minor because the block valve failure affected the reactor coolant system equipment and barrier performance attribute of the barrier integrity cornerstone. The inspectors evaluated the condition with the Phase 2 worksheet because the finding involved the reactor coolant system barrier. This finding is only of very low safety significance because the block valve inoperability did not significantly contribute to an increase in core damage frequency. The licensee placed this issue in their corrective action program as CAR 200306563.
Inspection Report# : 2003006(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective actions following an unanalyzed condition.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation was related to inadequate corrective actions taken following identification of an unanalyzed condition (control room ventilation envelope door open) which resulted in the postulated postaccident control room dose limits to be exceeded. The licensee's corrective actions failed to prevent recurrence of the condition.
This finding was greater than minor because it was associated with the integrity of the control room envelope. Because this finding involved the degradation of barrier integrity, the finding was evaluated using the significance determination process for at-power situations. The inspectors concluded that the finding was only of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2003005(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure of containment radiation monitors to meet Technical Specifications operability requirements.
The inspectors identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, associated with the licensee's failure to assure that applicable regulatory requirements and the design basis for the containment radiation gas monitors were correctly translated into Calculation GT-13 and, ultimately, the radiation monitor setpoint. This deficiency resulted in the containment gaseous channel becoming incapable of performing the design bases function to detect a one gallon per minute reactor coolant system leak within one hour in accordance with the licensee's commitment to Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems."
This finding was greater than minor because the containment gas channel radiation monitor was not capable of performing the design basis function for an extended period of time. The inoperability of the radiation monitor resulted in potential impact on reactor safety and adversely affected the reactor coolant leakage performance attribute of the barrier integrity reactor safety cornerstone. The finding was only of very low safety significance because other methods of reactor coolant system leak detection were available to the licensee. The unavailability of the gaseous channel leak detection function did not contribute to an increase in core damage sequences when evaluated using the significance 07/14/2004
 
1Q/2004 Inspection Findings - Callaway                                                                                                Page 4 of 5 determination process Phase 2 worksheets.
Inspection Report# : 2003005(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to barricade a high radiation area.
The inspectors identified a non-cited violation of Technical Specification 5.7.1 because the licensee failed to barricade a high radiation area to prevent inadvertent entry. Specifically, on October 21, 2003, while performing independent radiation measurements, the inspectors identified a high radiation area on the 2031-foot elevation of the radwaste building that was not enclosed by a barricade. Radiation dose rates around a demineralizer sample panel drain tank were as high as 140 millirems per hour at 30 centimeters from the surface penetrated by the radiation.
The finding is in the licensee's corrective action program as CAR 200307676.
This finding was greater than minor because inadequate controls of high radiation areas affect the licensee's ability to ensure adequate protection of worker health and safety from exposure to radiation and affected the cornerstone attribute/exposure control. Because the finding involved the potential for workers to receive significant unplanned, unintended dose as a result of conditions contrary to technical specification requirements, the inspector used the Occupational Radiation Safety Significance Determination Process described in Manual Chapter 0609, Appendix C, to analyze the significance of the finding. The inspector determined that a substantial potential for overexposure did not exist; therefore, the finding had very low significance.
Inspection Report# : 2003006(pdf)
Public Radiation Safety Significance:        Jul 02, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Dose rates on the external surface of a package in excess of DOT limits.
The licensee failed to maintain contact dose rates to 200 millirems per hour or less on a package transported in an open, exclusive use shipment, in violation of 49 CFR 173.441(b)(1).
This self-revealing, noncited violation was greater than minor because the finding is associated with one of the Public Radiation Safety Cornerstone attributes (transportation packaging) and the finding affects the associated cornerstone objective (to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain). The finding was related to an occurrence in the licensee's radioactive material transportation program that was contrary to Department of Transportation regulations and, therefore, was processed through the Public Radiation Safety Significance Determination Process. The finding is of very low safety significance because it involved a radiation dose limit (200 millirems per hour) that was exceeded, but the dose rate (300 millirems per hour) did not exceed the limit by more than two times and it was not accessible to the public.
Inspection Report# : 2003009(pdf)
Physical Protection Miscellaneous 07/14/2004
 
1Q/2004 Inspection Findings - Callaway                                                                                                  Page 5 of 5 Significance: N/A Jun 06, 2003 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems On the basis of the sample selected for review, the team concluded that in general, problems were adequately identified, evaluated, and corrected. The team identified a number of examples pertaining to the failure to promptly identify and correct conditions adverse to quality.
One long-standing issue involving a failure to promptly identify and correct voided conditions affecting both trains of the containment spray system suction piping following abnormal system response during surveillance testing on multiple occasions dating back to 1995 was identified by the team. Problem identification and resolution issues have affected Callaway historically and corrective actions have been put in place to improve performance. The team noted that engineering products reviewed effectively supported the corrective action process, were technically adequate, and provided sufficient justification to support operability for degraded conditions evaluated.
Inspection Report# : 2003010(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999.
In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : May 05, 2004 07/14/2004
 
2Q/2004 Inspection Findings - Callaway                                                                                                            Page 1 of 6 Callaway 2Q/2004 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Reactor trip during power ascension A self-revealing finding and a noncited violation of Technical Specification 5.4.1, "Procedures," was identified after an operator error resulted in an unplanned reactor trip. The operator's action to open the main feedwater regulating valves, before the plant was stable and at the prescribed power level, was the direct cause of the reactor trip.
This finding is greater than minor because the reactor trip was a transient initiator affecting the initiating events cornerstone. The operator's failure to follow the procedure was a performance deficiency which affected the human performance attribute of the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance (Green), because the condition did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation of equipment functions, and did not increase the likelihood of a fire or internal/external flood. The licensee placed the issue into the corrective action program as CAR 200401167.
Inspection Report# : 2004002(pdf)
Significance:        Mar 24, 2004 Identified By: Self Disclosing Item Type: FIN Finding Loss of the TDAFW pump during a transient A self-revealing finding was identified after the unplanned loss of the turbine- driven auxiliary feedwater pump during a plant transient. After a reactor trip, an operator improperly secured the turbine-driven auxiliary feedwater pump, which lead to an overspeed trip.
This finding was greater than minor because the loss of the turbine-driven feedwater pump affected the availability/reliability objective of the mitigating system equipment performance cornerstone. The inspectors concluded that this finding was only of very low safety significance because: it was not a design or qualification deficiency, it did not represent the actual loss of the safety function of a system, it did not represent the actual loss of the safety function of a single train for greater than its Technical Specification allowed outage time, it did not represent the loss of a non-Technical Specification related train (designated as risk significant per 10 CFR 50.65 a(4)) for greater than 24 hours, and it did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. The licensee's placed the issue into the corrective action program as CAR 200401167.
Inspection Report# : 2004002(pdf)
Significance:        Mar 24, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Safety injection due to operator error A self-revealing finding and a noncited violation of Technical Specification 5.4.1, "Procedures," was identified after an operator error resulted in an unplanned safety injection and main steamline isolation. The operator failed to place pressurizer pressure control in automatic during plant heatup operations. Pressurizer pressure exceeded the Permissive P-11 setpoint, while the main steamline pressure was still below the safety injection setpoint.
This finding is greater than minor because the safety injection was a transient initiator contributor affecting the initiating events cornerstone. The operator's failure to follow the procedure was a performance deficiency which affected the human performance attribute of the initiating events cornerstone. The inspectors concluded that this finding is of very low safety significance because the condition did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation of equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004002(pdf)
Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding The failure of a licensed operator to follow a procedure resulted in an unplanned plant transient.
An unplanned plant transient resulted from the failure of an operator to follow a written procedure. The transient occurred after the unexpected loss of all plant service cooling water and all but one of the condenser circulating water pumps. Cooling water was lost after an operator inadvertently opened the feeder breaker supplying power to the pumps.
 
2Q/2004 Inspection Findings - Callaway                                                                                                            Page 2 of 6 This finding is greater than minor because the operator error affected the human performance attribute of the initiating events cornerstone. The inspectors determined that the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood. The finding was similar to Example 4.b in MC 0612, Appendix E and was entered into the licensee's corrective action program as Callaway Action Request (CAR) 200308178.
Inspection Report# : 2003006(pdf)
Mitigating Systems Significance:        Mar 24, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate smoke alarm response procedure for control room supply.
The alarm response procedure for responding to smoke in the control room outside supply duct was inadequate because it did not direct operators to isolate outside air makeup upon receipt of the alarm. This alarm would not cause an automatic isolation of the control room, so operators must recognize the condition and take manual action to prevent losing control room habitability. Failure to have a procedure, required by Technical Specification 5.4.1.a and Regulatory Guide 1.33, that provided appropriate response actions for abnormal or alarm conditions was a violation. This issue was entered into the licensee's corrective action program under Callaway Action Request 200306977.
This issue was more than minor because failure to isolate the control room ventilation could lead to unnecessary evacuation, which would result in a plant transient and disabling much of the mitigation equipment that would otherwise be available. This issue was of very low safety significance because the frequency of the specific fire scenario necessary to cause an unnecessary control room evacuation was determined to be very small.
Inspection Report# : 2004002(pdf)
Significance:        Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Required Compensatory Measures When CREVIS Operation Rendered ESF Switchgear Room Halon System Inoperable The licensee did not recognize that the halon system protecting both engineered safety feature switchgear rooms was rendered inoperable and, therefore, failed to take the required compensatory action when the control room emergency ventilation and isolation system was in operation. Two ventilation dampers in parallel through the common fire wall between these rooms open when this system starts. The team identified that these dampers do not automatically shut when the halon system actuates. The halon system would not be capable of reaching the required concentration to suppress a fire because halon would be allowed to escape under these conditions. License Condition 2.C.(5)(c) requires that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Standardized Nuclear Unit Power Plant System Final Safety Analysis Report. Updated Final Safety Analysis Report, Table 9.5.1-2, "Halon Systems," requires that when this halon system is inoperable, the licensee shall establish a continuous fire watch with backup fire suppression capability in the affected area. Contrary to this, on numerous occasions throughout the operating life of the plant, the team found that the licensee had failed to post a continuous fire watch whenever the vital switchgear room halon system was rendered inoperable due to testing of the control room ventilation system. This violation of License Condition 2.C.(5)(c) will be treated as a noncited violation, consistent with Section VI.A of the Enforcement Policy. This issue was in the licensee's corrective action program under Callaway Action Request 200307189.
This finding was greater than minor because it involved the potential degradation of a fire protection feature protecting the electrical distribution equipment powering both trains of mitigating systems. This finding is of very low safety significance because the fire ignition frequency in the rooms affected is low, the remaining fire detection and suppression capability are unaffected, and sufficient accident mitigation equipment was available.
Inspection Report# : 2003007(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions following an EDG rocker arm lube oil valve mispositioning.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation was related to inadequate corrective actions taken following an emergency diesel generator rocker arm lube oil valve mispositioning. The licensee's corrective actions were not adequate to prevent recurrence.
This finding was greater than minor because it could reasonably be viewed as a precursor to a significant event and if left uncorrected, would become a more significant safety concern. This finding was of very low safety significance because the condition was not a design or qualification deficiency, did not represent the actual loss of a safety function of a system, did not represent the actual loss of a safety function of a single train for greater than its Technical Specification allowed outage time, did not represent the loss of a non-Technical Specification related train for greater than 24 hours, or did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Inspection Report# : 2003005(pdf)
 
2Q/2004 Inspection Findings - Callaway                                                                                                          Page 3 of 6 Barrier Integrity Significance:        Apr 12, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Work Instructions Resulted in the Failure of Residual Heat Removal Pump Seal.
Green. A self-revealing finding and noncited violation of Technical Specification 5.4.1, "Procedures," was identified after maintenance resulted in the failure of a residual heat removal pump seal during shutdown cooling operations. The licensee's maintenance work instructions were not adequate to ensure the mechanical seal matting ring surface was fully seated when replaced on March 31, 2004. The seal failed on April 11 after about 36 hours of operation.
This finding was greater than minor because it affected the barrier integrity cornerstone attribute of procedure quality, as related to maintenance procedures affecting the functionality of containment. The failed seal provided a containment leakage path for 7 gallons per minute reactor coolant. The inspectors evaluated the finding using the significance determination process for at-power situations because the issue involved the potential degradation of containment barrier integrity during power operations prior to the reactor shutdown on April 10. The finding was only of very low safety significance because the condition did not represent an actual open pathway in the physical integrity of reactor containment during power operation, was not an actual reduction of the atmospheric pressure control function of the reactor containment, and did not represent a degradation of a the control room auxiliary building or spent fuel pool barrier function. The licenses placed the issue into the corrective action program as CAR 200402749.
Inspection Report# : 2004003(pdf)
Significance:        Apr 02, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for implementation of TS 5.5.2.
The team identified a noncited violation of Technical Specification 5.4.1(e) for failure to establish an adequate procedure for evaluating emergency core cooling system leakage outside of containment as required by Technical Specification 5.5.2.
This finding was more than minor since it represented a programmatic weakness which, if left uncorrected could become a more significant safety concern. This finding screened as Green, very low safety significance, during the SDP Phase 1 analysis, because it only represented a degradation of the radiological barrier function provided for the control room and auxiliary building.
Inspection Report# : 2004006(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate postmaintenance test of a pressurizer power operated relief block valve.
The inspectors identified a finding and noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control." This finding is related to inadequate testing of the pressurizer power operated relief valve (PORV) block valve following modifications to the actuator circuit. The testing failed to detect that the valve actuator had failed.
This finding is greater than minor because the block valve failure affected the reactor coolant system equipment and barrier performance attribute of the barrier integrity cornerstone. The inspectors evaluated the condition with the Phase 2 worksheet because the finding involved the reactor coolant system barrier. The finding was only of very low safety significance because the block valve failure did not significantly contribute to an increase in core damage frequency. The licensee placed this issue in their corrective action program as CAR 200306563.
Inspection Report# : 2003006(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate incorporation of design information into work instructions lead to the failure of a pressurizer block valve.
The inspectors identified a finding and NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This finding is related to inadequate incorporation of design information into the work instructions for modifications to a pressurizer PORV block valve actuator circuit. The inadequate work instructions resulted in the failure of the valve actuator following return to service.
This finding is greater than minor because the block valve failure affected the reactor coolant system equipment and barrier performance attribute of the barrier integrity cornerstone. The inspectors evaluated the condition with the Phase 2 worksheet because the finding involved the reactor coolant system barrier. This finding is only of very low safety significance because the block valve inoperability did not significantly contribute to an increase in core damage frequency. The licensee placed this issue in their corrective action program as CAR 200306563.
Inspection Report# : 2003006(pdf)
Significance:        Sep 20, 2003
 
2Q/2004 Inspection Findings - Callaway                                                                                                        Page 4 of 6 Identified By: NRC Item Type: NCV NonCited Violation Ineffective actions following an unanalyzed condition.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation was related to inadequate corrective actions taken following identification of an unanalyzed condition (control room ventilation envelope door open) which resulted in the postulated postaccident control room dose limits to be exceeded. The licensee's corrective actions failed to prevent recurrence of the condition.
This finding was greater than minor because it was associated with the integrity of the control room envelope. Because this finding involved the degradation of barrier integrity, the finding was evaluated using the significance determination process for at-power situations. The inspectors concluded that the finding was only of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2003005(pdf)
Significance:        Sep 20, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure of containment radiation monitors to meet Technical Specifications operability requirements.
The inspectors identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, associated with the licensee's failure to assure that applicable regulatory requirements and the design basis for the containment radiation gas monitors were correctly translated into Calculation GT-13 and, ultimately, the radiation monitor setpoint. This deficiency resulted in the containment gaseous channel becoming incapable of performing the design bases function to detect a one gallon per minute reactor coolant system leak within one hour in accordance with the licensee's commitment to Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems."
This finding was greater than minor because the containment gas channel radiation monitor was not capable of performing the design basis function for an extended period of time. The inoperability of the radiation monitor resulted in potential impact on reactor safety and adversely affected the reactor coolant leakage performance attribute of the barrier integrity reactor safety cornerstone. The finding was only of very low safety significance because other methods of reactor coolant system leak detection were available to the licensee. The unavailability of the gaseous channel leak detection function did not contribute to an increase in core damage sequences when evaluated using the significance determination process Phase 2 worksheets.
Inspection Report# : 2003005(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Jun 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Operational Control Resulted in an Unexpected High Radiation Field.
Green. A self-revealing finding and NCV of Technical Specification 5.4.1 was identified after three plant workers were exposed to an unplanned high radiation area. The event was the result of inadequate operational control of the in-core system. The exposure occurred when a reactor engineer removed two in-core detectors from the core after control room personnel authorized a reactor building entry. The procedure used by the reactor engineer to operate the in-core system was not appropriate to the circumstances.
The inspectors used the occupational radiation safety determination processes to analyze the significance of the finding. This finding was greater than minor because it affected the programs and process attribute of the occupational radiation safety cornerstone. The use of the inappropriate procedure could have resulted in unplanned or unintended dose which could have been significantly greater as a result of a single, minor, alteration of the circumstances. The inspectors concluded the issue was of very low safety significance because the inspection finding was not related to as low as is reasonably achievable, did not involve an overexposure, and there was no substantial potential for overexposure. The licensee entered this issue into the corrective action program as Callaway Action Request 200402640. This issue was determined to have crosscutting aspects regarding human performance.
Inspection Report# : 2004003(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to barricade a high radiation area.
The inspectors identified a non-cited violation of Technical Specification 5.7.1 because the licensee failed to barricade a high radiation area to prevent inadvertent entry. Specifically, on October 21, 2003, while performing independent radiation measurements, the inspectors identified a high radiation
 
2Q/2004 Inspection Findings - Callaway                                                                                                        Page 5 of 6 area on the 2031-foot elevation of the radwaste building that was not enclosed by a barricade. Radiation dose rates around a demineralizer sample panel drain tank were as high as 140 millirems per hour at 30 centimeters from the surface penetrated by the radiation. The finding is in the licensee's corrective action program as CAR 200307676.
This finding was greater than minor because inadequate controls of high radiation areas affect the licensee's ability to ensure adequate protection of worker health and safety from exposure to radiation and affected the cornerstone attribute/exposure control. Because the finding involved the potential for workers to receive significant unplanned, unintended dose as a result of conditions contrary to technical specification requirements, the inspector used the Occupational Radiation Safety Significance Determination Process described in Manual Chapter 0609, Appendix C, to analyze the significance of the finding. The inspector determined that a substantial potential for overexposure did not exist; therefore, the finding had very low significance.
Inspection Report# : 2003006(pdf)
Public Radiation Safety Significance:        Jul 02, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Dose rates on the external surface of a package in excess of DOT limits.
The licensee failed to maintain contact dose rates to 200 millirems per hour or less on a package transported in an open, exclusive use shipment, in violation of 49 CFR 173.441(b)(1).
This self-revealing, noncited violation was greater than minor because the finding is associated with one of the Public Radiation Safety Cornerstone attributes (transportation packaging) and the finding affects the associated cornerstone objective (to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain). The finding was related to an occurrence in the licensee's radioactive material transportation program that was contrary to Department of Transportation regulations and, therefore, was processed through the Public Radiation Safety Significance Determination Process. The finding is of very low safety significance because it involved a radiation dose limit (200 millirems per hour) that was exceeded, but the dose rate (300 millirems per hour) did not exceed the limit by more than two times and it was not accessible to the public.
Inspection Report# : 2003009(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:        Apr 02, 2004 Identified By: NRC Item Type: FIN Finding Identification and resolution of problems.
The team reviewed approximately 105 corrective action documents, 28 self-assessments and audits, and numerous procedures, industry information, and other documents. The team determined that there was a general improvement in implementation of the corrective action program; thresholds for identifying issues remained appropriately low, and in most cases, corrective actions were adequate to address conditions adverse to quality. However, in some instances, improper prioritization or the lack of a rigorous evaluation of problems continued to challenge the licensee. The team also concluded that a safety conscious work environment exists at Callaway, however some negative comments received during interviews indicated that efforts to improve in this area have not been completely effective.
Inspection Report# : 2004006(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999.
 
2Q/2004 Inspection Findings - Callaway                                                                                                          Page 6 of 6 In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : September 08, 2004
 
3Q/2004 Inspection Findings - Callaway                                                                                                    Page 1 of 6 Callaway 3Q/2004 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Safety injection due to operator error A self-revealing finding and a noncited violation of Technical Specification 5.4.1, "Procedures," was identified after an operator error resulted in an unplanned safety injection and main steamline isolation. The operator failed to place pressurizer pressure control in automatic during plant heatup operations. Pressurizer pressure exceeded the Permissive P-11 setpoint, while the main steamline pressure was still below the safety injection setpoint.
This finding is greater than minor because the safety injection was a transient initiator contributor affecting the initiating events cornerstone.
The operator's failure to follow the procedure was a performance deficiency which affected the human performance attribute of the initiating events cornerstone. The inspectors concluded that this finding is of very low safety significance because the condition did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation of equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004002(pdf)
Significance:        Mar 24, 2004 Identified By: Self Disclosing Item Type: FIN Finding Loss of the TDAFW pump during a transient A self-revealing finding was identified after the unplanned loss of the turbine- driven auxiliary feedwater pump during a plant transient. After a reactor trip, an operator improperly secured the turbine-driven auxiliary feedwater pump, which lead to an overspeed trip.
This finding was greater than minor because the loss of the turbine-driven feedwater pump affected the availability/reliability objective of the mitigating system equipment performance cornerstone. The inspectors concluded that this finding was only of very low safety significance because: it was not a design or qualification deficiency, it did not represent the actual loss of the safety function of a system, it did not represent the actual loss of the safety function of a single train for greater than its Technical Specification allowed outage time, it did not represent the loss of a non-Technical Specification related train (designated as risk significant per 10 CFR 50.65 a(4)) for greater than 24 hours, and it did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. The licensee's placed the issue into the corrective action program as CAR 200401167.
Inspection Report# : 2004002(pdf)
Significance:        Mar 24, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Reactor trip during power ascension A self-revealing finding and a noncited violation of Technical Specification 5.4.1, "Procedures," was identified after an operator error resulted in an unplanned reactor trip. The operator's action to open the main feedwater regulating valves, before the plant was stable and at the prescribed power level, was the direct cause of the reactor trip.
This finding is greater than minor because the reactor trip was a transient initiator affecting the initiating events cornerstone. The operator's failure to follow the procedure was a performance deficiency which affected the human performance attribute of the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance (Green), because the condition did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation of equipment functions, and did not increase the likelihood of a fire or internal/external flood. The licensee placed the issue into the corrective action program as CAR 200401167.
Inspection Report# : 2004002(pdf)
Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding The failure of a licensed operator to follow a procedure resulted in an unplanned plant transient.
 
3Q/2004 Inspection Findings - Callaway                                                                                                  Page 2 of 6 An unplanned plant transient resulted from the failure of an operator to follow a written procedure. The transient occurred after the unexpected loss of all plant service cooling water and all but one of the condenser circulating water pumps. Cooling water was lost after an operator inadvertently opened the feeder breaker supplying power to the pumps.
This finding is greater than minor because the operator error affected the human performance attribute of the initiating events cornerstone. The inspectors determined that the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood. The finding was similar to Example 4.b in MC 0612, Appendix E and was entered into the licensee's corrective action program as Callaway Action Request (CAR) 200308178.
Inspection Report# : 2003006(pdf)
Mitigating Systems Significance:        Sep 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate selection and suitability review of installation of lead radiation shield blankets in containment.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to perform an adequate selection and suitability review prior to installing 132 lead radiation shield blankets in containment. The licensee did not address the effect that blankets may have on safety related equipment during accident conditions. During an accident, some of the blanket coverings/coatings may deteriorate into foreign material and be transported to the containment sump. Once at the sump, this foreign material may challenge emergency core cooling system recirculation function by reducing the available net positive suction head to the residual heat removal and containment spray pumps.
The finding is greater than minor because it affected the cornerstone objective to ensure availability and reliability of the containment sump.
This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per GL 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. The licensee placed this issue in their corrective action program as CAR 200404836.
Inspection Report# : 2004004(pdf)
Significance:        Mar 24, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate smoke alarm response procedure for control room supply.
The alarm response procedure for responding to smoke in the control room outside supply duct was inadequate because it did not direct operators to isolate outside air makeup upon receipt of the alarm. This alarm would not cause an automatic isolation of the control room, so operators must recognize the condition and take manual action to prevent losing control room habitability. Failure to have a procedure, required by Technical Specification 5.4.1.a and Regulatory Guide 1.33, that provided appropriate response actions for abnormal or alarm conditions was a violation. This issue was entered into the licensee's corrective action program under Callaway Action Request 200306977.
This issue was more than minor because failure to isolate the control room ventilation could lead to unnecessary evacuation, which would result in a plant transient and disabling much of the mitigation equipment that would otherwise be available. This issue was of very low safety significance because the frequency of the specific fire scenario necessary to cause an unnecessary control room evacuation was determined to be very small.
Inspection Report# : 2004002(pdf)
Significance:        Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Required Compensatory Measures When CREVIS Operation Rendered ESF Switchgear Room Halon System Inoperable The licensee did not recognize that the halon system protecting both engineered safety feature switchgear rooms was rendered inoperable and, therefore, failed to take the required compensatory action when the control room emergency ventilation and isolation system was in operation.
Two ventilation dampers in parallel through the common fire wall between these rooms open when this system starts. The team identified that these dampers do not automatically shut when the halon system actuates. The halon system would not be capable of reaching the required concentration to suppress a fire because halon would be allowed to escape under these conditions. License Condition 2.C.(5)(c) requires that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Standardized Nuclear Unit Power Plant System Final Safety Analysis Report. Updated Final Safety Analysis Report, Table 9.5.1-2, "Halon Systems," requires that when this halon system is inoperable, the licensee shall establish a continuous fire watch with backup fire suppression capability in the affected
 
3Q/2004 Inspection Findings - Callaway                                                                                                Page 3 of 6 area. Contrary to this, on numerous occasions throughout the operating life of the plant, the team found that the licensee had failed to post a continuous fire watch whenever the vital switchgear room halon system was rendered inoperable due to testing of the control room ventilation system. This violation of License Condition 2.C.(5)(c) will be treated as a noncited violation, consistent with Section VI.A of the Enforcement Policy. This issue was in the licensee's corrective action program under Callaway Action Request 200307189.
This finding was greater than minor because it involved the potential degradation of a fire protection feature protecting the electrical distribution equipment powering both trains of mitigating systems. This finding is of very low safety significance because the fire ignition frequency in the rooms affected is low, the remaining fire detection and suppression capability are unaffected, and sufficient accident mitigation equipment was available.
Inspection Report# : 2003007(pdf)
Barrier Integrity Significance:        Sep 28, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to test automatic recirculation control valves recirculation isolation feature.
A noncited violation of 10 CFR Part 50, Appendix B, Criteria XI, "Test Control," was identified for the failure to establish a test procedure with appropriate acceptance criteria to verify the proper operation of the auxiliary feedwater system automatic recirculation control valves. This issue was entered into the corrective action program as Callaway Action Request 200407321.
The finding is greater than minor because it affected the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding is associated with the cornerstone attribute of procedure quality. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process,"
this finding is determined to be of every low safety significance because there was no actual loss of a safety function.
Inspection Report# : 2004008(pdf)
Significance:        Apr 12, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Work Instructions Resulted in the Failure of Residual Heat Removal Pump Seal.
Green. A self-revealing finding and noncited violation of Technical Specification 5.4.1, "Procedures," was identified after maintenance resulted in the failure of a residual heat removal pump seal during shutdown cooling operations. The licensee's maintenance work instructions were not adequate to ensure the mechanical seal matting ring surface was fully seated when replaced on March 31, 2004. The seal failed on April 11 after about 36 hours of operation.
This finding was greater than minor because it affected the barrier integrity cornerstone attribute of procedure quality, as related to maintenance procedures affecting the functionality of containment. The failed seal provided a containment leakage path for 7 gallons per minute reactor coolant. The inspectors evaluated the finding using the significance determination process for at-power situations because the issue involved the potential degradation of containment barrier integrity during power operations prior to the reactor shutdown on April 10. The finding was only of very low safety significance because the condition did not represent an actual open pathway in the physical integrity of reactor containment during power operation, was not an actual reduction of the atmospheric pressure control function of the reactor containment, and did not represent a degradation of a the control room auxiliary building or spent fuel pool barrier function. The licenses placed the issue into the corrective action program as CAR 200402749.
Inspection Report# : 2004003(pdf)
Significance:        Apr 02, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for implementation of TS 5.5.2.
The team identified a noncited violation of Technical Specification 5.4.1(e) for failure to establish an adequate procedure for evaluating emergency core cooling system leakage outside of containment as required by Technical Specification 5.5.2.
This finding was more than minor since it represented a programmatic weakness which, if left uncorrected could become a more significant safety concern. This finding screened as Green, very low safety significance, during the SDP Phase 1 analysis, because it only represented a degradation of the radiological barrier function provided for the control room and auxiliary building.
Inspection Report# : 2004006(pdf)
 
3Q/2004 Inspection Findings - Callaway                                                                                              Page 4 of 6 Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate incorporation of design information into work instructions lead to the failure of a pressurizer block valve.
The inspectors identified a finding and NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This finding is related to inadequate incorporation of design information into the work instructions for modifications to a pressurizer PORV block valve actuator circuit.
The inadequate work instructions resulted in the failure of the valve actuator following return to service.
This finding is greater than minor because the block valve failure affected the reactor coolant system equipment and barrier performance attribute of the barrier integrity cornerstone. The inspectors evaluated the condition with the Phase 2 worksheet because the finding involved the reactor coolant system barrier. This finding is only of very low safety significance because the block valve inoperability did not significantly contribute to an increase in core damage frequency. The licensee placed this issue in their corrective action program as CAR 200306563.
Inspection Report# : 2003006(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Inadequate postmaintenance test of a pressurizer power operated relief block valve.
The inspectors identified a finding and noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control." This finding is related to inadequate testing of the pressurizer power operated relief valve (PORV) block valve following modifications to the actuator circuit. The testing failed to detect that the valve actuator had failed.
This finding is greater than minor because the block valve failure affected the reactor coolant system equipment and barrier performance attribute of the barrier integrity cornerstone. The inspectors evaluated the condition with the Phase 2 worksheet because the finding involved the reactor coolant system barrier. The finding was only of very low safety significance because the block valve failure did not significantly contribute to an increase in core damage frequency. The licensee placed this issue in their corrective action program as CAR 200306563.
Inspection Report# : 2003006(pdf)
Emergency Preparedness Significance:        Sep 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to classify and declare an unusual event following a fire in the protected area.
The inspectors identified a noncited violation of 10CFR50.54(q), 10CFR50.47(b)(4), and Section IV.B of Appendix E of 10CFR Part 50, which involved the failure to correctly classify an UE in accordance with the emergency plan and implementing procedures. The operations crew did not activate the emergency plan for a fire in the proteted area, adjacent to the control building, which lasted longer than 15 minutes from verification. This finding has human performance crosscutting aspects in that the licensee failed to properly apply event evaluation criteria.
This finding is more than minor because it affected the response organization performance attribute of the emergency preparedness cornerstone due to failure to properly recognize plant conditions commensurate with an UE classification. This finding was of very log safety significance, because it did not meet any higher level emergency plan and implementing procedure notification requirements. The licensee placed the issue into the corrective action program as Callaway Action Request 200407284.
Inspection Report# : 2004004(pdf)
Occupational Radiation Safety Significance:        Jun 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Operational Control Resulted in an Unexpected High Radiation Field.
Green. A self-revealing finding and NCV of Technical Specification 5.4.1 was identified after three plant workers were exposed to an unplanned high radiation area. The event was the result of inadequate operational control of the in-core system. The exposure occurred when a reactor engineer removed two in-core detectors from the core after control room personnel authorized a reactor building entry. The procedure
 
3Q/2004 Inspection Findings - Callaway                                                                                                Page 5 of 6 used by the reactor engineer to operate the in-core system was not appropriate to the circumstances.
The inspectors used the occupational radiation safety determination processes to analyze the significance of the finding. This finding was greater than minor because it affected the programs and process attribute of the occupational radiation safety cornerstone. The use of the inappropriate procedure could have resulted in unplanned or unintended dose which could have been significantly greater as a result of a single, minor, alteration of the circumstances. The inspectors concluded the issue was of very low safety significance because the inspection finding was not related to as low as is reasonably achievable, did not involve an overexposure, and there was no substantial potential for overexposure.
The licensee entered this issue into the corrective action program as Callaway Action Request 200402640. This issue was determined to have crosscutting aspects regarding human performance.
Inspection Report# : 2004003(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to barricade a high radiation area.
The inspectors identified a non-cited violation of Technical Specification 5.7.1 because the licensee failed to barricade a high radiation area to prevent inadvertent entry. Specifically, on October 21, 2003, while performing independent radiation measurements, the inspectors identified a high radiation area on the 2031-foot elevation of the radwaste building that was not enclosed by a barricade. Radiation dose rates around a demineralizer sample panel drain tank were as high as 140 millirems per hour at 30 centimeters from the surface penetrated by the radiation.
The finding is in the licensee's corrective action program as CAR 200307676.
This finding was greater than minor because inadequate controls of high radiation areas affect the licensee's ability to ensure adequate protection of worker health and safety from exposure to radiation and affected the cornerstone attribute/exposure control. Because the finding involved the potential for workers to receive significant unplanned, unintended dose as a result of conditions contrary to technical specification requirements, the inspector used the Occupational Radiation Safety Significance Determination Process described in Manual Chapter 0609, Appendix C, to analyze the significance of the finding. The inspector determined that a substantial potential for overexposure did not exist; therefore, the finding had very low significance.
Inspection Report# : 2003006(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:        Apr 02, 2004 Identified By: NRC Item Type: FIN Finding Identification and resolution of problems.
The team reviewed approximately 105 corrective action documents, 28 self-assessments and audits, and numerous procedures, industry information, and other documents. The team determined that there was a general improvement in implementation of the corrective action program; thresholds for identifying issues remained appropriately low, and in most cases, corrective actions were adequate to address conditions adverse to quality. However, in some instances, improper prioritization or the lack of a rigorous evaluation of problems continued to challenge the licensee. The team also concluded that a safety conscious work environment exists at Callaway, however some negative comments received during interviews indicated that efforts to improve in this area have not been completely effective.
Inspection Report# : 2004006(pdf)
Significance: SL-III May 14, 2001 Identified By: NRC Item Type: VIO Violation Discrimination against a security officer and a training instructor for having engaged in protected activity 10 CFR 50.7(a) prohibits discrimination by a Commission licensee against an employee for engaging in certain protected activities. On October
 
3Q/2004 Inspection Findings - Callaway                                                                                                  Page 6 of 6 27, 1999, the security officer and the training instructor identified to the Wackenhut Corporation a violation of NRC requirements at the Callaway Nuclear Plant. Based at least in part on this protected activity, the Wackenhut Corporation unfavorably terminated the security officer's employment for lack of trustworthiness and gave a written reprimand to the training instructor on November 19, 1999.
In consideration of the severity of the actions taken against the former security officer and the training instructor, the level of management involved in the adverse action, and the nature of contractor/licensee relationships, this violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III (EA-01-005, dated May 14, 2001).
Inspection Report# : 2001003(pdf)
Last modified : December 29, 2004
 
4Q/2004 Inspection Findings - Callaway                                                                                                  Page 1 of 6 Callaway 4Q/2004 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding Operator Error Resulted in a Steam Generator Chemistry Excursion.
A self-revealing finding was identified after an operator error resulted in an unplanned secondary side chemistry excursion and a steam generator blowdown isolation. An operator failed to maintain minimum cooling tower blowdown flow during an effluent release of steam generator blowdown demineralizer flush water to the environment. The reduction in flow resulted in the isolation of the release and pressurization of the steam generator blowdown flush line. The pressurized line resulted in the transfer of flush water to the main condenser and caused steam generator chemistry to exceed the Action Level 2 threshold. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance (personnel).
This finding is greater than minor because the chemistry excursion had an impact on the equipment performance attribute of the initiating events objective cornerstone. The inspectors determined that this finding is of very low safety significance because the chemistry excursion did not add to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004005(pdf)
Significance: N/A Nov 08, 2004 Identified By: NRC Item Type: FIN Finding Supplemental Inspection for a White performance indicator in the initiating events cornerstone.
The NRC conducted a supplemental inspection to assess the licensee's evaluation of conditions associated with a White performance indicator in the initiating events cornerstone. Three unplanned reactor trips resulted in the unplanned scrams per 7,000 critical hours performance indicator to cross the threshold from Green to White during the second quarter of 2004. The inspector concluded that the licensee's problem identification, root cause, extent-of-condition evaluations, and corrective actions for the three reactor trips were adequate. Two of the reactor trips were caused by main generator supervisory relay failures. The third reactor trip was caused by a reactor operator's failure to follow the power ascension procedure. Several of the root causes contributing to the third reactor trip have been long-standing station problems. The inspector identified weaknesses in the licensee's root cause determination and corrective actions related to the third reactor trip. The inspector did not identify any common attributes linking the three reactor trips from a risk perspective.
Inspection Report# : 2004009(pdf)
Significance:        Mar 24, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Safety injection due to operator error A self-revealing finding and a noncited violation of Technical Specification 5.4.1, "Procedures," was identified after an operator error resulted in an unplanned safety injection and main steamline isolation. The operator failed to place pressurizer pressure control in automatic during plant heatup operations. Pressurizer pressure exceeded the Permissive P-11 setpoint, while the main steamline pressure was still below the safety injection setpoint.
This finding is greater than minor because the safety injection was a transient initiator contributor affecting the initiating events cornerstone.
The operator's failure to follow the procedure was a performance deficiency which affected the human performance attribute of the initiating events cornerstone. The inspectors concluded that this finding is of very low safety significance because the condition did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation of equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004002(pdf)
Significance:        Mar 24, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Reactor trip during power ascension A self-revealing finding and a noncited violation of Technical Specification 5.4.1, "Procedures," was identified after an operator error resulted in an unplanned reactor trip. The operator's action to open the main feedwater regulating valves, before the plant was stable and at the
 
4Q/2004 Inspection Findings - Callaway                                                                                                  Page 2 of 6 prescribed power level, was the direct cause of the reactor trip.
This finding is greater than minor because the reactor trip was a transient initiator affecting the initiating events cornerstone. The operator's failure to follow the procedure was a performance deficiency which affected the human performance attribute of the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance (Green), because the condition did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to a loss of mitigation of equipment functions, and did not increase the likelihood of a fire or internal/external flood. The licensee placed the issue into the corrective action program as CAR 200401167.
Inspection Report# : 2004002(pdf)
Mitigating Systems Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain the Integrity of an Auxiliary Building Fire Door The inspectors identified a noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," after the licensee failed to maintain the integrity of an auxiliary building fire door. The inspectors identified that the fire door would not provide the rated fire confinement function because of a broken latch. The door provided the 3-hour fire barrier between auxiliary building fire Areas A-19 and A-20. The licensee had several prior opportunities to identify the degraded fire door. The plant security procedure required plant security officers to verify that the fire door was properly latched during each patrol. Several security patrols passed through the door each shift. This finding had crosscutting aspects related to human performance (personnel) in that the plant procedure regarding verification of fire doors was not followed.
This finding is greater than minor because the fire door was associated with the mitigating system cornerstone attribute to provide protection against external factors. The inspectors concluded that the degraded door was a fire confinement finding with a high degradation rating due to the broken latch. This finding is of very low safety significance because the degraded door did not separate unique potential fire damage targets and that the door would have provided at least 20 minutes fire endurance protection. The inspectors also concluded that no fixed or in-situ fire ignition sources or combustible or flammable materials were positioned such that the degraded door would have been subject to direct flame impingement.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Required Fire Watch.
The inspectors identified a noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," after a plant fire occurred when the licensee failed to establish a required fire watch. A welder ignited a fire on the communication corridor roof. The fire burned through the roof and ignited the ceiling below. The licensee had not established a fire watch inside the room. The plant fire brigade responded and extinguished the fire. The fire brigade left the area without establishing a re-flash fire watch. About 55 minutes later, an equipment operator returned to the room and identified that the fire had reignited. The plant fire brigade responded again and extinguished the re-flash fire.
This finding is greater than minor because the mitigating systems cornerstone attribute providing protection against external factors was affected. This finding had an adverse affect on the licensee's fire protection defense-in-depth strategies related to fire detection, manual suppression, and fire brigade effectiveness. The inspectors concluded that the lack of a fire watch degraded the licensee's early fire suppression capability and resulted in the fire prevention finding with a high degradation rating. The inspectors determined that this finding is of very low significance because the fire ignition source could not have caused ignition of secondary combustible fuels and was not close enough to sufficient surrounding combustibles to cause damage consistent with any of the plant fire damage scenarios.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Postmaintenance Test Failed to Identify Degraded Turbine Driven Auxiliary Feedwater Pump Bearing Cooling Following Maintenance.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," after postmaintenance testing was not adequate to identify degraded turbine-driven auxiliary feedwater pump bearing cooling following maintenance. The licensee completed an overhaul of the turbine, performed a postmaintenance test, and returned the system to service. Twenty-four days later, the licensee observed elevated inboard turbine bearing temperatures during a surveillance test. The elevated temperatures were caused by an obstruction in the lube oil cooler. The lube oil filter had been improperly installed during the overhaul and allowed particulate material to bypass the filter. The inspectors identified that an elevated bearing temperature also occurred during the earlier postmaintenance test. However, the licensee did not
 
4Q/2004 Inspection Findings - Callaway                                                                                                  Page 3 of 6 monitor bearing temperatures during the test nor was postmaintenance testing performed for a sufficient duration to allow the turbine to reach normal operating temperatures. This finding had crosscutting aspects regarding human performance (personnel) for failure to adequately test the turbine-driven auxiliary feedwater pump following maintenance, and problem identification in that indications were present during an earlier test that should have alerted the licensee to the condition.
This finding is greater than minor because, if left uncorrected the condition would become a more significant safety concern. This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2004005(pdf)
Significance:        Sep 28, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to test automatic recirculation control valves recirculation isolation feature.
A noncited violation of 10 CFR Part 50, Appendix B, Criteria XI, "Test Control," was identified for the failure to establish a test procedure with appropriate acceptance criteria to verify the proper operation of the auxiliary feedwater system automatic recirculation control valves. This issue was entered into the corrective action program as Callaway Action Request 200407321.
The finding is greater than minor because it affected the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding is associated with the cornerstone attribute of procedure quality. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process,"
this finding is determined to be of every low safety significance because there was no actual loss of a safety function.
Inspection Report# : 2004008(pdf)
Significance:        Sep 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate selection and suitability review of installation of lead radiation shield blankets in containment.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to perform an adequate selection and suitability review prior to installing 132 lead radiation shield blankets in containment. The licensee did not address the effect that blankets may have on safety related equipment during accident conditions. During an accident, some of the blanket coverings/coatings may deteriorate into foreign material and be transported to the containment sump. Once at the sump, this foreign material may challenge emergency core cooling system recirculation function by reducing the available net positive suction head to the residual heat removal and containment spray pumps.
The finding is greater than minor because it affected the cornerstone objective to ensure availability and reliability of the containment sump.
This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per GL 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. The licensee placed this issue in their corrective action program as CAR 200404836.
Inspection Report# : 2004004(pdf)
Significance:        Mar 24, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate smoke alarm response procedure for control room supply.
The alarm response procedure for responding to smoke in the control room outside supply duct was inadequate because it did not direct operators to isolate outside air makeup upon receipt of the alarm. This alarm would not cause an automatic isolation of the control room, so operators must recognize the condition and take manual action to prevent losing control room habitability. Failure to have a procedure, required by Technical Specification 5.4.1.a and Regulatory Guide 1.33, that provided appropriate response actions for abnormal or alarm conditions was a violation. This issue was entered into the licensee's corrective action program under Callaway Action Request 200306977.
This issue was more than minor because failure to isolate the control room ventilation could lead to unnecessary evacuation, which would result in a plant transient and disabling much of the mitigation equipment that would otherwise be available. This issue was of very low safety significance because the frequency of the specific fire scenario necessary to cause an unnecessary control room evacuation was determined to be very small.
Inspection Report# : 2004002(pdf)
Significance:        Mar 24, 2004 Identified By: Self Disclosing
 
4Q/2004 Inspection Findings - Callaway                                                                                                    Page 4 of 6 Item Type: FIN Finding Loss of the TDAFW pump during a transient A self-revealing finding was identified after the unplanned loss of the turbine- driven auxiliary feedwater pump during a plant transient. After a reactor trip, an operator improperly secured the turbine-driven auxiliary feedwater pump, which lead to an overspeed trip.
This finding was greater than minor because the loss of the turbine-driven feedwater pump affected the availability/reliability objective of the mitigating system equipment performance cornerstone. The inspectors concluded that this finding was only of very low safety significance because: it was not a design or qualification deficiency, it did not represent the actual loss of the safety function of a system, it did not represent the actual loss of the safety function of a single train for greater than its Technical Specification allowed outage time, it did not represent the loss of a non-Technical Specification related train (designated as risk significant per 10 CFR 50.65 a(4)) for greater than 24 hours, and it did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. The licensee's placed the issue into the corrective action program as CAR 200401167.
Inspection Report# : 2004002(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Feedwater Isolation Valve Post Modification Deficiencies.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," after the licensee failed to correct deficiencies identified during post modification testing of the feedwater isolation valve actuators. The post modification test revealed that the feedwater isolation valves would not meet the Mode 3 closure times described in the licensing bases. The licensee dispositioned the deficiency without adequately correcting the deficiencies. The licensee had a second opportunity to identify the inadequate corrective actions when the Independent Technical Review Team assessed the post modification test results. The Independent Technical Review Team assessment was not effective in identifying the inadequate corrective actions. This finding has crosscutting aspects regarding failure to implement adequate corrective actions.
This finding is greater than minor because the failure of the feedwater isolation valves to meet closures times affected the barrier integrity cornerstone design control attribute to maintain the functionality of the fuel cladding, following a cooldown event, and to limit post accident containment pressure by isolating feedwater to the faulted steam generator. This finding is only of very low safety significance because the condition did not represent a degradation of the barrier function of the control room, auxiliary building, or spent fuel pool, nor did this finding represent an actual open pathway in the physical integrity of the containment, nor affect the atmospheric pressure control or hydrogen control functions of containment.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation An Operator Error Resulted in an Unplanned Transfer of Water from Spent Fuel Pool.
A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an operator error resulted in the unplanned transfer of 3600 gallons of water from the spent fuel pool. The operating procedure required the operator to shutdown refueling water storage tank recirculation before placing fuel pool cleanup in service. The operator failed to shutdown the recirculation lineup resulting in the unplanned spent fuel pool water loss. The operating crew recognized the decreasing spent fuel pool level and secured the recirculation after about 3600 gallons had been transferred.
This finding is greater than minor because if left uncorrected it would have become a more significant safety concern. The inspectors determined that this finding is only of very low significance because the condition only represented a degradation of the radiological barrier function provided by the spent fuel pool.
Inspection Report# : 2004005(pdf)
Significance:        Apr 12, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Work Instructions Resulted in the Failure of Residual Heat Removal Pump Seal.
Green. A self-revealing finding and noncited violation of Technical Specification 5.4.1, "Procedures," was identified after maintenance resulted in the failure of a residual heat removal pump seal during shutdown cooling operations. The licensee's maintenance work instructions were not adequate to ensure the mechanical seal matting ring surface was fully seated when replaced on March 31, 2004. The seal failed on April 11 after about 36 hours of operation.
 
4Q/2004 Inspection Findings - Callaway                                                                                                Page 5 of 6 This finding was greater than minor because it affected the barrier integrity cornerstone attribute of procedure quality, as related to maintenance procedures affecting the functionality of containment. The failed seal provided a containment leakage path for 7 gallons per minute reactor coolant. The inspectors evaluated the finding using the significance determination process for at-power situations because the issue involved the potential degradation of containment barrier integrity during power operations prior to the reactor shutdown on April 10. The finding was only of very low safety significance because the condition did not represent an actual open pathway in the physical integrity of reactor containment during power operation, was not an actual reduction of the atmospheric pressure control function of the reactor containment, and did not represent a degradation of a the control room auxiliary building or spent fuel pool barrier function. The licenses placed the issue into the corrective action program as CAR 200402749.
Inspection Report# : 2004003(pdf)
Significance:        Apr 02, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for implementation of TS 5.5.2.
The team identified a noncited violation of Technical Specification 5.4.1(e) for failure to establish an adequate procedure for evaluating emergency core cooling system leakage outside of containment as required by Technical Specification 5.5.2.
This finding was more than minor since it represented a programmatic weakness which, if left uncorrected could become a more significant safety concern. This finding screened as Green, very low safety significance, during the SDP Phase 1 analysis, because it only represented a degradation of the radiological barrier function provided for the control room and auxiliary building.
Inspection Report# : 2004006(pdf)
Emergency Preparedness Significance:        Sep 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to classify and declare an unusual event following a fire in the protected area.
The inspectors identified a noncited violation of 10CFR50.54(q), 10CFR50.47(b)(4), and Section IV.B of Appendix E of 10CFR Part 50, which involved the failure to correctly classify an UE in accordance with the emergency plan and implementing procedures. The operations crew did not activate the emergency plan for a fire in the proteted area, adjacent to the control building, which lasted longer than 15 minutes from verification. This finding has human performance crosscutting aspects in that the licensee failed to properly apply event evaluation criteria.
This finding is more than minor because it affected the response organization performance attribute of the emergency preparedness cornerstone due to failure to properly recognize plant conditions commensurate with an UE classification. This finding was of very log safety significance, because it did not meet any higher level emergency plan and implementing procedure notification requirements. The licensee placed the issue into the corrective action program as Callaway Action Request 200407284.
Inspection Report# : 2004004(pdf)
Occupational Radiation Safety Significance:        Jun 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Operational Control Resulted in an Unexpected High Radiation Field.
Green. A self-revealing finding and NCV of Technical Specification 5.4.1 was identified after three plant workers were exposed to an unplanned high radiation area. The event was the result of inadequate operational control of the in-core system. The exposure occurred when a reactor engineer removed two in-core detectors from the core after control room personnel authorized a reactor building entry. The procedure used by the reactor engineer to operate the in-core system was not appropriate to the circumstances.
The inspectors used the occupational radiation safety determination processes to analyze the significance of the finding. This finding was greater than minor because it affected the programs and process attribute of the occupational radiation safety cornerstone. The use of the inappropriate procedure could have resulted in unplanned or unintended dose which could have been significantly greater as a result of a single, minor, alteration of the circumstances. The inspectors concluded the issue was of very low safety significance because the inspection finding was not related to as low as is reasonably achievable, did not involve an overexposure, and there was no substantial potential for overexposure.
The licensee entered this issue into the corrective action program as Callaway Action Request 200402640. This issue was determined to have
 
4Q/2004 Inspection Findings - Callaway                                                                                            Page 6 of 6 crosscutting aspects regarding human performance.
Inspection Report# : 2004003(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:        Apr 02, 2004 Identified By: NRC Item Type: FIN Finding Identification and resolution of problems.
The team reviewed approximately 105 corrective action documents, 28 self-assessments and audits, and numerous procedures, industry information, and other documents. The team determined that there was a general improvement in implementation of the corrective action program; thresholds for identifying issues remained appropriately low, and in most cases, corrective actions were adequate to address conditions adverse to quality. However, in some instances, improper prioritization or the lack of a rigorous evaluation of problems continued to challenge the licensee. The team also concluded that a safety conscious work environment exists at Callaway, however some negative comments received during interviews indicated that efforts to improve in this area have not been completely effective.
Inspection Report# : 2004006(pdf)
Last modified : March 09, 2005
 
1Q/2005 Inspection Findings - Callaway                                                                                                  Page 1 of 6 Callaway 1Q/2005 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2005 Identified By: Self Disclosing Item Type: FIN Finding Unplanned reactor trip due to ineffective use of industry OE during a maintenance activity.
A self-revealing finding was identified after an unplanned reactor trip resulted from the licensee's ineffective use of industry operating experience. The plant tripped from low steam generator level after a feedwater regulating valve closed. The regulating valve closed after a control power supply shorted during a maintenance activity. The power supply shorted because the maintenance workers had used an inadequate work instruction. A similar event occurred at the Beaver Valley Nuclear Plant during June 2003. The licensee failed to effectively use the operating experience when planning and performing the maintenance activity. The licensee's failure to properly revise an incorrect work package before proceeding with the work activity, a poor prejob brief, and organizational time pressures also contributed to the event.
Additionally, the licensee's evaluation of the event identified contributing causes as root causes, and did not take into account the programmatic issues to include operating experience reviews into work instruction development procedures. This finding had crosscutting aspects regarding human performance, and problem identification and resolution in that the evaluation of root versus contributing causes was deficient.
This finding was more than minor because the procedural adequacy attribute of the initiating events cornerstone objective was affected. The inspectors concluded the reactor trip is a transient initiator, affecting the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment functions.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding Operator Error Resulted in a Steam Generator Chemistry Excursion.
A self-revealing finding was identified after an operator error resulted in an unplanned secondary side chemistry excursion and a steam generator blowdown isolation. An operator failed to maintain minimum cooling tower blowdown flow during an effluent release of steam generator blowdown demineralizer flush water to the environment. The reduction in flow resulted in the isolation of the release and pressurization of the steam generator blowdown flush line. The pressurized line resulted in the transfer of flush water to the main condenser and caused steam generator chemistry to exceed the Action Level 2 threshold. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance (personnel).
This finding is greater than minor because the chemistry excursion had an impact on the equipment performance attribute of the initiating events objective cornerstone. The inspectors determined that this finding is of very low safety significance because the chemistry excursion did not add to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004005(pdf)
Significance: N/A Nov 08, 2004 Identified By: NRC Item Type: FIN Finding Supplemental Inspection for a White performance indicator in the initiating events cornerstone.
The NRC conducted a supplemental inspection to assess the licensee's evaluation of conditions associated with a White performance indicator in the initiating events cornerstone. Three unplanned reactor trips resulted in the unplanned scrams per 7,000 critical hours performance indicator to cross the threshold from Green to White during the second quarter of 2004. The inspector concluded that the licensee's problem identification, root cause, extent-of-condition evaluations, and corrective actions for the three reactor trips were adequate. Two of the reactor trips were caused by main generator supervisory relay failures. The third reactor trip was caused by a reactor operator's failure to follow the power ascension procedure. Several of the root causes contributing to the third reactor trip have been long-standing station problems. The inspector identified weaknesses in the licensee's root cause determination and corrective actions related to the third reactor trip. The inspector did not identify any common attributes linking the three reactor trips from a risk perspective.
Inspection Report# : 2004009(pdf)
 
1Q/2005 Inspection Findings - Callaway                                                                                                  Page 2 of 6 Mitigating Systems Significance:        Mar 24, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain the minimum number of fire brigade members on site.
An NRC identified noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee failed to maintain the minimum number of fire brigade members on site. The inspectors identified that the licensee did not maintain minimum fire brigade staffing. The licensee was required to maintain at least five fire brigade members on site at all times. Between January 24 and February 9, 2005, the outside equipment operator was assigned to the fire brigade 68 percent of the time. However, the outside equipment operator spent about 80 percent of the shift outside of the protected area, including attending equipment at the river pumping station, located eight miles from the site. The inspectors concluded that full fire brigade staffing would have been delayed about 20 to 30 minutes if the activation occurred while the equipment operator was performing outside duties. This finding had crosscutting aspects regarding human performance in that full fire brigade staffing was not ensured. This finding also had crosscutting aspects regarding problem identification and resolution in that the issue was not properly evaluated following documentation in the corrective action program twice.
This finding is greater than minor because the reactor safety mitigating systems cornerstone objective attribute to provide protection against external factors was affected. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," does not address fire brigade performance deficiencies. Regional management review concluded this finding was of very low safety significance because it affected the fire prevention and administrative controls category and represented only a short duration degradation in fire brigade staffing.
Inspection Report# : 2005002(pdf)
Significance:        Mar 24, 2005 Identified By: Self Disclosing Item Type: NCV NonCited Violation Unplanned loss of water fire supression due to an inadequate testing procedure.
A self revealing noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee inadvertently isolated all plant fire water suppression from the reactor, auxiliary, control, and turbine buildings during surveillance testing. The isolation resulted in the unplanned loss of all fire water to the reactor, auxiliary, control, and turbine buildings. The isolation occurred due an inadequate surveillance testing procedure. The licensee identified the isolation of the fire loops after about 15 minutes. The licensee reestablished the fire water suppression system after about 1.5 hours. This finding had crosscutting aspects regarding human performance in that the procedure used was inadequate.
The finding is greater than minor because the unplanned isolation of fire water was associated with the "Protection Against External Factors,"
attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure availability of systems designed to respond to initiating events. The inspectors used Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the condition had an adverse affect on fire defense-in-depth strategies. The senior reactor analyst evaluated the finding based on a bounding calculation for each fire area affected by the loss of fire water in the plant. The analyst concluded a plant-wide fire mitigation probability of 4.3 x 10-6 over the 2-hour exposure period. The analyst assumed that the maximum Conditional Core Damage Probability for any fire area was bounded by probability used to assess fires requiring control room evacuation (CCDP=0.1). The maximum resulting core damage probability from internal fires over the 2-hour period was the product of the plant-wide fire mitigation probability and 0.1. This bounded the risk of the finding resulting in no greater increase in core damage frequency than 4.3 x 10-7. The analyst concluded that a systematic search and assessment effort was beyond the intended scope of the fire protection significance determination process. Therefore, in accordance with NRC Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Section 05.04.c, regional management reviewed this finding and determined that it was of very low risk significance.
Inspection Report# : 2005002(pdf)
Significance:        Mar 24, 2005 Identified By: NRC Item Type: NCV NonCited Violation Ineffective cause determination and corrective actions to prevent recurrence of ECCS pipe voiding.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, after the licensee's cause determination and corrective actions were ineffective to prevent recurrence of safety injection pump discharge pipe voiding. Plant Technical Specifications required the licensee to verify that the emergency core cooling system piping was full of water every 31 days. The licensee established a 20 percent maximum void fraction as the acceptance limit for the safety injection pump hot leg injection discharge piping. On seven occasions during the past 2 years the surveillance acceptance criteria was not met. This finding had crosscutting aspects regarding problem identification and resolution in that the licensee's actions to determine the cause of the repeated surveillance failures and to implement corrective actions were not effective in preventing recurrence of the condition.
This finding is greater than minor because voiding in emergency core cooling system piping affected the reactor mitigating systems cornerstone and the equipment performance attribute to ensure availability of systems that respond to prevent core damage. This finding was only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic
 
1Q/2005 Inspection Findings - Callaway                                                                                                  Page 3 of 6 Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Required Fire Watch.
The inspectors identified a noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," after a plant fire occurred when the licensee failed to establish a required fire watch. A welder ignited a fire on the communication corridor roof. The fire burned through the roof and ignited the ceiling below. The licensee had not established a fire watch inside the room. The plant fire brigade responded and extinguished the fire. The fire brigade left the area without establishing a re-flash fire watch. About 55 minutes later, an equipment operator returned to the room and identified that the fire had reignited. The plant fire brigade responded again and extinguished the re-flash fire.
This finding is greater than minor because the mitigating systems cornerstone attribute providing protection against external factors was affected. This finding had an adverse affect on the licensee's fire protection defense-in-depth strategies related to fire detection, manual suppression, and fire brigade effectiveness. The inspectors concluded that the lack of a fire watch degraded the licensee's early fire suppression capability and resulted in the fire prevention finding with a high degradation rating. The inspectors determined that this finding is of very low significance because the fire ignition source could not have caused ignition of secondary combustible fuels and was not close enough to sufficient surrounding combustibles to cause damage consistent with any of the plant fire damage scenarios.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain the Integrity of an Auxiliary Building Fire Door The inspectors identified a noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," after the licensee failed to maintain the integrity of an auxiliary building fire door. The inspectors identified that the fire door would not provide the rated fire confinement function because of a broken latch. The door provided the 3-hour fire barrier between auxiliary building fire Areas A-19 and A-20. The licensee had several prior opportunities to identify the degraded fire door. The plant security procedure required plant security officers to verify that the fire door was properly latched during each patrol. Several security patrols passed through the door each shift. This finding had crosscutting aspects related to human performance (personnel) in that the plant procedure regarding verification of fire doors was not followed.
This finding is greater than minor because the fire door was associated with the mitigating system cornerstone attribute to provide protection against external factors. The inspectors concluded that the degraded door was a fire confinement finding with a high degradation rating due to the broken latch. This finding is of very low safety significance because the degraded door did not separate unique potential fire damage targets and that the door would have provided at least 20 minutes fire endurance protection. The inspectors also concluded that no fixed or in-situ fire ignition sources or combustible or flammable materials were positioned such that the degraded door would have been subject to direct flame impingement.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Postmaintenance Test Failed to Identify Degraded Turbine Driven Auxiliary Feedwater Pump Bearing Cooling Following Maintenance.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," after postmaintenance testing was not adequate to identify degraded turbine-driven auxiliary feedwater pump bearing cooling following maintenance. The licensee completed an overhaul of the turbine, performed a postmaintenance test, and returned the system to service. Twenty-four days later, the licensee observed elevated inboard turbine bearing temperatures during a surveillance test. The elevated temperatures were caused by an obstruction in the lube oil cooler. The lube oil filter had been improperly installed during the overhaul and allowed particulate material to bypass the filter. The inspectors identified that an elevated bearing temperature also occurred during the earlier postmaintenance test. However, the licensee did not monitor bearing temperatures during the test nor was postmaintenance testing performed for a sufficient duration to allow the turbine to reach normal operating temperatures. This finding had crosscutting aspects regarding human performance (personnel) for failure to adequately test the turbine-driven auxiliary feedwater pump following maintenance, and problem identification in that indications were present during an earlier test that should have alerted the licensee to the condition.
This finding is greater than minor because, if left uncorrected the condition would become a more significant safety concern. This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2004005(pdf)
 
1Q/2005 Inspection Findings - Callaway                                                                                                  Page 4 of 6 Significance:        Sep 28, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to test automatic recirculation control valves recirculation isolation feature.
A noncited violation of 10 CFR Part 50, Appendix B, Criteria XI, "Test Control," was identified for the failure to establish a test procedure with appropriate acceptance criteria to verify the proper operation of the auxiliary feedwater system automatic recirculation control valves. This issue was entered into the corrective action program as Callaway Action Request 200407321.
The finding is greater than minor because it affected the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding is associated with the cornerstone attribute of procedure quality. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process,"
this finding is determined to be of every low safety significance because there was no actual loss of a safety function.
Inspection Report# : 2004008(pdf)
Significance:        Sep 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate selection and suitability review of installation of lead radiation shield blankets in containment.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to perform an adequate selection and suitability review prior to installing 132 lead radiation shield blankets in containment. The licensee did not address the effect that blankets may have on safety related equipment during accident conditions. During an accident, some of the blanket coverings/coatings may deteriorate into foreign material and be transported to the containment sump. Once at the sump, this foreign material may challenge emergency core cooling system recirculation function by reducing the available net positive suction head to the residual heat removal and containment spray pumps.
The finding is greater than minor because it affected the cornerstone objective to ensure availability and reliability of the containment sump.
This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per GL 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. The licensee placed this issue in their corrective action program as CAR 200404836.
Inspection Report# : 2004004(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Feedwater Isolation Valve Post Modification Deficiencies.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," after the licensee failed to correct deficiencies identified during post modification testing of the feedwater isolation valve actuators. The post modification test revealed that the feedwater isolation valves would not meet the Mode 3 closure times described in the licensing bases. The licensee dispositioned the deficiency without adequately correcting the deficiencies. The licensee had a second opportunity to identify the inadequate corrective actions when the Independent Technical Review Team assessed the post modification test results. The Independent Technical Review Team assessment was not effective in identifying the inadequate corrective actions. This finding has crosscutting aspects regarding failure to implement adequate corrective actions.
This finding is greater than minor because the failure of the feedwater isolation valves to meet closures times affected the barrier integrity cornerstone design control attribute to maintain the functionality of the fuel cladding, following a cooldown event, and to limit post accident containment pressure by isolating feedwater to the faulted steam generator. This finding is only of very low safety significance because the condition did not represent a degradation of the barrier function of the control room, auxiliary building, or spent fuel pool, nor did this finding represent an actual open pathway in the physical integrity of the containment, nor affect the atmospheric pressure control or hydrogen control functions of containment.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation An Operator Error Resulted in an Unplanned Transfer of Water from Spent Fuel Pool.
 
1Q/2005 Inspection Findings - Callaway                                                                                                Page 5 of 6 A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an operator error resulted in the unplanned transfer of 3600 gallons of water from the spent fuel pool. The operating procedure required the operator to shutdown refueling water storage tank recirculation before placing fuel pool cleanup in service. The operator failed to shutdown the recirculation lineup resulting in the unplanned spent fuel pool water loss. The operating crew recognized the decreasing spent fuel pool level and secured the recirculation after about 3600 gallons had been transferred.
This finding is greater than minor because if left uncorrected it would have become a more significant safety concern. The inspectors determined that this finding is only of very low significance because the condition only represented a degradation of the radiological barrier function provided by the spent fuel pool.
Inspection Report# : 2004005(pdf)
Significance:        Apr 12, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Work Instructions Resulted in the Failure of Residual Heat Removal Pump Seal.
Green. A self-revealing finding and noncited violation of Technical Specification 5.4.1, "Procedures," was identified after maintenance resulted in the failure of a residual heat removal pump seal during shutdown cooling operations. The licensee's maintenance work instructions were not adequate to ensure the mechanical seal matting ring surface was fully seated when replaced on March 31, 2004. The seal failed on April 11 after about 36 hours of operation.
This finding was greater than minor because it affected the barrier integrity cornerstone attribute of procedure quality, as related to maintenance procedures affecting the functionality of containment. The failed seal provided a containment leakage path for 7 gallons per minute reactor coolant. The inspectors evaluated the finding using the significance determination process for at-power situations because the issue involved the potential degradation of containment barrier integrity during power operations prior to the reactor shutdown on April 10. The finding was only of very low safety significance because the condition did not represent an actual open pathway in the physical integrity of reactor containment during power operation, was not an actual reduction of the atmospheric pressure control function of the reactor containment, and did not represent a degradation of a the control room auxiliary building or spent fuel pool barrier function. The licenses placed the issue into the corrective action program as CAR 200402749.
Inspection Report# : 2004003(pdf)
Significance:        Apr 02, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for implementation of TS 5.5.2.
The team identified a noncited violation of Technical Specification 5.4.1(e) for failure to establish an adequate procedure for evaluating emergency core cooling system leakage outside of containment as required by Technical Specification 5.5.2.
This finding was more than minor since it represented a programmatic weakness which, if left uncorrected could become a more significant safety concern. This finding screened as Green, very low safety significance, during the SDP Phase 1 analysis, because it only represented a degradation of the radiological barrier function provided for the control room and auxiliary building.
Inspection Report# : 2004006(pdf)
Emergency Preparedness Significance:        Sep 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to classify and declare an unusual event following a fire in the protected area.
The inspectors identified a noncited violation of 10CFR50.54(q), 10CFR50.47(b)(4), and Section IV.B of Appendix E of 10CFR Part 50, which involved the failure to correctly classify an UE in accordance with the emergency plan and implementing procedures. The operations crew did not activate the emergency plan for a fire in the proteted area, adjacent to the control building, which lasted longer than 15 minutes from verification. This finding has human performance crosscutting aspects in that the licensee failed to properly apply event evaluation criteria.
This finding is more than minor because it affected the response organization performance attribute of the emergency preparedness cornerstone due to failure to properly recognize plant conditions commensurate with an UE classification. This finding was of very log safety significance, because it did not meet any higher level emergency plan and implementing procedure notification requirements. The licensee placed the issue into the corrective action program as Callaway Action Request 200407284.
Inspection Report# : 2004004(pdf)
 
1Q/2005 Inspection Findings - Callaway                                                                                              Page 6 of 6 Occupational Radiation Safety Significance:        Jun 23, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Operational Control Resulted in an Unexpected High Radiation Field.
Green. A self-revealing finding and NCV of Technical Specification 5.4.1 was identified after three plant workers were exposed to an unplanned high radiation area. The event was the result of inadequate operational control of the in-core system. The exposure occurred when a reactor engineer removed two in-core detectors from the core after control room personnel authorized a reactor building entry. The procedure used by the reactor engineer to operate the in-core system was not appropriate to the circumstances.
The inspectors used the occupational radiation safety determination processes to analyze the significance of the finding. This finding was greater than minor because it affected the programs and process attribute of the occupational radiation safety cornerstone. The use of the inappropriate procedure could have resulted in unplanned or unintended dose which could have been significantly greater as a result of a single, minor, alteration of the circumstances. The inspectors concluded the issue was of very low safety significance because the inspection finding was not related to as low as is reasonably achievable, did not involve an overexposure, and there was no substantial potential for overexposure.
The licensee entered this issue into the corrective action program as Callaway Action Request 200402640. This issue was determined to have crosscutting aspects regarding human performance.
Inspection Report# : 2004003(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:        Apr 02, 2004 Identified By: NRC Item Type: FIN Finding Identification and resolution of problems.
The team reviewed approximately 105 corrective action documents, 28 self-assessments and audits, and numerous procedures, industry information, and other documents. The team determined that there was a general improvement in implementation of the corrective action program; thresholds for identifying issues remained appropriately low, and in most cases, corrective actions were adequate to address conditions adverse to quality. However, in some instances, improper prioritization or the lack of a rigorous evaluation of problems continued to challenge the licensee. The team also concluded that a safety conscious work environment exists at Callaway, however some negative comments received during interviews indicated that efforts to improve in this area have not been completely effective.
Inspection Report# : 2004006(pdf)
Last modified : June 17, 2005
 
2Q/2005 Inspection Findings - Callaway                                                                                                  Page 1 of 6 Callaway 2Q/2005 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2005 Identified By: Self Disclosing Item Type: NCV NonCited Violation Unplanned auxiliary feedwater actuation due to use of an inadequate general operating procedure for troubleshooting.
A self-revealing noncited violation of Technical Specification 5.4.1.a was identified after an unplanned auxiliary feedwater actuation and reactor trip signal occurred while shutdown due to an inadequate general operating procedure and poor crew decision making.
This finding is greater than minor because the procedural adequacy attribute of the initiating events cornerstone objective is affected. The inspectors concluded the auxiliary feedwater actuation and reactor trip signal was a transient initiator, affecting the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment functions.
Inspection Report# : 2005003(pdf)
Significance:        Mar 24, 2005 Identified By: Self Disclosing Item Type: FIN Finding Unplanned reactor trip due to ineffective use of industry OE during a maintenance activity.
A self-revealing finding was identified after an unplanned reactor trip resulted from the licensee's ineffective use of industry operating experience. The plant tripped from low steam generator level after a feedwater regulating valve closed. The regulating valve closed after a control power supply shorted during a maintenance activity. The power supply shorted because the maintenance workers had used an inadequate work instruction. A similar event occurred at the Beaver Valley Nuclear Plant during June 2003. The licensee failed to effectively use the operating experience when planning and performing the maintenance activity. The licensee's failure to properly revise an incorrect work package before proceeding with the work activity, a poor prejob brief, and organizational time pressures also contributed to the event.
Additionally, the licensee's evaluation of the event identified contributing causes as root causes, and did not take into account the programmatic issues to include operating experience reviews into work instruction development procedures. This finding had crosscutting aspects regarding human performance, and problem identification and resolution in that the evaluation of root versus contributing causes was deficient.
This finding was more than minor because the procedural adequacy attribute of the initiating events cornerstone objective was affected. The inspectors concluded the reactor trip is a transient initiator, affecting the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment functions.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding Operator Error Resulted in a Steam Generator Chemistry Excursion.
A self-revealing finding was identified after an operator error resulted in an unplanned secondary side chemistry excursion and a steam generator blowdown isolation. An operator failed to maintain minimum cooling tower blowdown flow during an effluent release of steam generator blowdown demineralizer flush water to the environment. The reduction in flow resulted in the isolation of the release and pressurization of the steam generator blowdown flush line. The pressurized line resulted in the transfer of flush water to the main condenser and caused steam generator chemistry to exceed the Action Level 2 threshold. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance (personnel).
This finding is greater than minor because the chemistry excursion had an impact on the equipment performance attribute of the initiating events objective cornerstone. The inspectors determined that this finding is of very low safety significance because the chemistry excursion did not add to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004005(pdf)
Significance: N/A Nov 08, 2004 Identified By: NRC Item Type: FIN Finding
 
2Q/2005 Inspection Findings - Callaway                                                                                                  Page 2 of 6 Supplemental Inspection for a White performance indicator in the initiating events cornerstone.
The NRC conducted a supplemental inspection to assess the licensee's evaluation of conditions associated with a White performance indicator in the initiating events cornerstone. Three unplanned reactor trips resulted in the unplanned scrams per 7,000 critical hours performance indicator to cross the threshold from Green to White during the second quarter of 2004. The inspector concluded that the licensee's problem identification, root cause, extent-of-condition evaluations, and corrective actions for the three reactor trips were adequate. Two of the reactor trips were caused by main generator supervisory relay failures. The third reactor trip was caused by a reactor operator's failure to follow the power ascension procedure. Several of the root causes contributing to the third reactor trip have been long-standing station problems. The inspector identified weaknesses in the licensee's root cause determination and corrective actions related to the third reactor trip. The inspector did not identify any common attributes linking the three reactor trips from a risk perspective.
Inspection Report# : 2004009(pdf)
Mitigating Systems Significance:        Jun 23, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain the integrity of a three-hour auxiliary building fire door.
A self-revealing noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee failed to maintain the integrity of an auxiliary building fire door that was required to provide a three-hour fire barrier.
This finding is greater than minor because the reactor safety mitigating systems cornerstone attribute to provide protection against external factors was affected. The inspectors used Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the degraded door is a fire barrier related to the licensee's fire protection defense-in-depth strategies. The licensee had several prior opportunities to self-identify the degraded door and previous corrective actions were not effective to prevent recurrence. The inspectors concluded that the condition was intermittent and thus had a low degradation rating. The inspectors concluded this finding is of very low safety significance because of the low degradation level.
Inspection Report# : 2005003(pdf)
Significance:        Mar 24, 2005 Identified By: Self Disclosing Item Type: NCV NonCited Violation Unplanned loss of water fire supression due to an inadequate testing procedure.
A self revealing noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee inadvertently isolated all plant fire water suppression from the reactor, auxiliary, control, and turbine buildings during surveillance testing. The isolation resulted in the unplanned loss of all fire water to the reactor, auxiliary, control, and turbine buildings. The isolation occurred due an inadequate surveillance testing procedure. The licensee identified the isolation of the fire loops after about 15 minutes. The licensee reestablished the fire water suppression system after about 1.5 hours. This finding had crosscutting aspects regarding human performance in that the procedure used was inadequate.
The finding is greater than minor because the unplanned isolation of fire water was associated with the "Protection Against External Factors,"
attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure availability of systems designed to respond to initiating events. The inspectors used Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the condition had an adverse affect on fire defense-in-depth strategies. The senior reactor analyst evaluated the finding based on a bounding calculation for each fire area affected by the loss of fire water in the plant. The analyst concluded a plant-wide fire mitigation probability of 4.3 x 10-6 over the 2-hour exposure period. The analyst assumed that the maximum Conditional Core Damage Probability for any fire area was bounded by probability used to assess fires requiring control room evacuation (CCDP=0.1). The maximum resulting core damage probability from internal fires over the 2-hour period was the product of the plant-wide fire mitigation probability and 0.1. This bounded the risk of the finding resulting in no greater increase in core damage frequency than 4.3 x 10-7. The analyst concluded that a systematic search and assessment effort was beyond the intended scope of the fire protection significance determination process. Therefore, in accordance with NRC Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Section 05.04.c, regional management reviewed this finding and determined that it was of very low risk significance.
Inspection Report# : 2005002(pdf)
Significance:        Mar 24, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain the minimum number of fire brigade members on site.
An NRC identified noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee failed to maintain the minimum number of fire brigade members on site. The inspectors identified that the licensee did not maintain minimum fire brigade staffing. The licensee was required to maintain at least five fire brigade members on site at all times. Between January 24 and February
 
2Q/2005 Inspection Findings - Callaway                                                                                                  Page 3 of 6 9, 2005, the outside equipment operator was assigned to the fire brigade 68 percent of the time. However, the outside equipment operator spent about 80 percent of the shift outside of the protected area, including attending equipment at the river pumping station, located eight miles from the site. The inspectors concluded that full fire brigade staffing would have been delayed about 20 to 30 minutes if the activation occurred while the equipment operator was performing outside duties. This finding had crosscutting aspects regarding human performance in that full fire brigade staffing was not ensured. This finding also had crosscutting aspects regarding problem identification and resolution in that the issue was not properly evaluated following documentation in the corrective action program twice.
This finding is greater than minor because the reactor safety mitigating systems cornerstone objective attribute to provide protection against external factors was affected. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," does not address fire brigade performance deficiencies. Regional management review concluded this finding was of very low safety significance because it affected the fire prevention and administrative controls category and represented only a short duration degradation in fire brigade staffing.
Inspection Report# : 2005002(pdf)
Significance:        Mar 24, 2005 Identified By: NRC Item Type: NCV NonCited Violation Ineffective cause determination and corrective actions to prevent recurrence of ECCS pipe voiding.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, after the licensee's cause determination and corrective actions were ineffective to prevent recurrence of safety injection pump discharge pipe voiding. Plant Technical Specifications required the licensee to verify that the emergency core cooling system piping was full of water every 31 days. The licensee established a 20 percent maximum void fraction as the acceptance limit for the safety injection pump hot leg injection discharge piping. On seven occasions during the past 2 years the surveillance acceptance criteria was not met. This finding had crosscutting aspects regarding problem identification and resolution in that the licensee's actions to determine the cause of the repeated surveillance failures and to implement corrective actions were not effective in preventing recurrence of the condition.
This finding is greater than minor because voiding in emergency core cooling system piping affected the reactor mitigating systems cornerstone and the equipment performance attribute to ensure availability of systems that respond to prevent core damage. This finding was only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Postmaintenance Test Failed to Identify Degraded Turbine Driven Auxiliary Feedwater Pump Bearing Cooling Following Maintenance.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," after postmaintenance testing was not adequate to identify degraded turbine-driven auxiliary feedwater pump bearing cooling following maintenance. The licensee completed an overhaul of the turbine, performed a postmaintenance test, and returned the system to service. Twenty-four days later, the licensee observed elevated inboard turbine bearing temperatures during a surveillance test. The elevated temperatures were caused by an obstruction in the lube oil cooler. The lube oil filter had been improperly installed during the overhaul and allowed particulate material to bypass the filter. The inspectors identified that an elevated bearing temperature also occurred during the earlier postmaintenance test. However, the licensee did not monitor bearing temperatures during the test nor was postmaintenance testing performed for a sufficient duration to allow the turbine to reach normal operating temperatures. This finding had crosscutting aspects regarding human performance (personnel) for failure to adequately test the turbine-driven auxiliary feedwater pump following maintenance, and problem identification in that indications were present during an earlier test that should have alerted the licensee to the condition.
This finding is greater than minor because, if left uncorrected the condition would become a more significant safety concern. This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Required Fire Watch.
The inspectors identified a noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," after a plant fire occurred when the licensee failed to establish a required fire watch. A welder ignited a fire on the communication corridor roof. The fire burned through the roof and ignited the ceiling below. The licensee had not established a fire watch inside the room. The plant fire brigade responded and extinguished the fire. The fire brigade left the area without establishing a re-flash fire watch. About 55 minutes later, an equipment operator returned to the room and identified that the fire had reignited. The plant fire brigade responded again and extinguished the re-flash fire.
 
2Q/2005 Inspection Findings - Callaway                                                                                                  Page 4 of 6 This finding is greater than minor because the mitigating systems cornerstone attribute providing protection against external factors was affected. This finding had an adverse affect on the licensee's fire protection defense-in-depth strategies related to fire detection, manual suppression, and fire brigade effectiveness. The inspectors concluded that the lack of a fire watch degraded the licensee's early fire suppression capability and resulted in the fire prevention finding with a high degradation rating. The inspectors determined that this finding is of very low significance because the fire ignition source could not have caused ignition of secondary combustible fuels and was not close enough to sufficient surrounding combustibles to cause damage consistent with any of the plant fire damage scenarios.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain the Integrity of an Auxiliary Building Fire Door The inspectors identified a noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," after the licensee failed to maintain the integrity of an auxiliary building fire door. The inspectors identified that the fire door would not provide the rated fire confinement function because of a broken latch. The door provided the 3-hour fire barrier between auxiliary building fire Areas A-19 and A-20. The licensee had several prior opportunities to identify the degraded fire door. The plant security procedure required plant security officers to verify that the fire door was properly latched during each patrol. Several security patrols passed through the door each shift. This finding had crosscutting aspects related to human performance (personnel) in that the plant procedure regarding verification of fire doors was not followed.
This finding is greater than minor because the fire door was associated with the mitigating system cornerstone attribute to provide protection against external factors. The inspectors concluded that the degraded door was a fire confinement finding with a high degradation rating due to the broken latch. This finding is of very low safety significance because the degraded door did not separate unique potential fire damage targets and that the door would have provided at least 20 minutes fire endurance protection. The inspectors also concluded that no fixed or in-situ fire ignition sources or combustible or flammable materials were positioned such that the degraded door would have been subject to direct flame impingement.
Inspection Report# : 2004005(pdf)
Significance:        Sep 28, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to test automatic recirculation control valves recirculation isolation feature.
A noncited violation of 10 CFR Part 50, Appendix B, Criteria XI, "Test Control," was identified for the failure to establish a test procedure with appropriate acceptance criteria to verify the proper operation of the auxiliary feedwater system automatic recirculation control valves. This issue was entered into the corrective action program as Callaway Action Request 200407321.
The finding is greater than minor because it affected the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding is associated with the cornerstone attribute of procedure quality. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process,"
this finding is determined to be of every low safety significance because there was no actual loss of a safety function.
Inspection Report# : 2004008(pdf)
Significance:        Sep 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate selection and suitability review of installation of lead radiation shield blankets in containment.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to perform an adequate selection and suitability review prior to installing 132 lead radiation shield blankets in containment. The licensee did not address the effect that blankets may have on safety related equipment during accident conditions. During an accident, some of the blanket coverings/coatings may deteriorate into foreign material and be transported to the containment sump. Once at the sump, this foreign material may challenge emergency core cooling system recirculation function by reducing the available net positive suction head to the residual heat removal and containment spray pumps.
The finding is greater than minor because it affected the cornerstone objective to ensure availability and reliability of the containment sump.
This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per GL 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. The licensee placed this issue in their corrective action program as CAR 200404836.
Inspection Report# : 2004004(pdf)
 
2Q/2005 Inspection Findings - Callaway                                                                                                Page 5 of 6 Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Feedwater Isolation Valve Post Modification Deficiencies.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," after the licensee failed to correct deficiencies identified during post modification testing of the feedwater isolation valve actuators. The post modification test revealed that the feedwater isolation valves would not meet the Mode 3 closure times described in the licensing bases. The licensee dispositioned the deficiency without adequately correcting the deficiencies. The licensee had a second opportunity to identify the inadequate corrective actions when the Independent Technical Review Team assessed the post modification test results. The Independent Technical Review Team assessment was not effective in identifying the inadequate corrective actions. This finding has crosscutting aspects regarding failure to implement adequate corrective actions.
This finding is greater than minor because the failure of the feedwater isolation valves to meet closures times affected the barrier integrity cornerstone design control attribute to maintain the functionality of the fuel cladding, following a cooldown event, and to limit post accident containment pressure by isolating feedwater to the faulted steam generator. This finding is only of very low safety significance because the condition did not represent a degradation of the barrier function of the control room, auxiliary building, or spent fuel pool, nor did this finding represent an actual open pathway in the physical integrity of the containment, nor affect the atmospheric pressure control or hydrogen control functions of containment.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation An Operator Error Resulted in an Unplanned Transfer of Water from Spent Fuel Pool.
A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an operator error resulted in the unplanned transfer of 3600 gallons of water from the spent fuel pool. The operating procedure required the operator to shutdown refueling water storage tank recirculation before placing fuel pool cleanup in service. The operator failed to shutdown the recirculation lineup resulting in the unplanned spent fuel pool water loss. The operating crew recognized the decreasing spent fuel pool level and secured the recirculation after about 3600 gallons had been transferred.
This finding is greater than minor because if left uncorrected it would have become a more significant safety concern. The inspectors determined that this finding is only of very low significance because the condition only represented a degradation of the radiological barrier function provided by the spent fuel pool.
Inspection Report# : 2004005(pdf)
Emergency Preparedness Significance:        Sep 23, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to classify and declare an unusual event following a fire in the protected area.
The inspectors identified a noncited violation of 10CFR50.54(q), 10CFR50.47(b)(4), and Section IV.B of Appendix E of 10CFR Part 50, which involved the failure to correctly classify an UE in accordance with the emergency plan and implementing procedures. The operations crew did not activate the emergency plan for a fire in the proteted area, adjacent to the control building, which lasted longer than 15 minutes from verification. This finding has human performance crosscutting aspects in that the licensee failed to properly apply event evaluation criteria.
This finding is more than minor because it affected the response organization performance attribute of the emergency preparedness cornerstone due to failure to properly recognize plant conditions commensurate with an UE classification. This finding was of very log safety significance, because it did not meet any higher level emergency plan and implementing procedure notification requirements. The licensee placed the issue into the corrective action program as Callaway Action Request 200407284.
Inspection Report# : 2004004(pdf)
Occupational Radiation Safety
 
2Q/2005 Inspection Findings - Callaway                  Page 6 of 6 Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : August 24, 2005
 
3Q/2005 Inspection Findings - Callaway                                                                                                  Page 1 of 6 Callaway 3Q/2005 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Unplanned auxiliary feedwater actuation due to use of an inadequate general operating procedure for troubleshooting.
A self-revealing noncited violation of Technical Specification 5.4.1.a was identified after an unplanned auxiliary feedwater actuation and reactor trip signal occurred while shutdown due to an inadequate general operating procedure and poor crew decision making.
This finding is greater than minor because the procedural adequacy attribute of the initiating events cornerstone objective is affected. The inspectors concluded the auxiliary feedwater actuation and reactor trip signal was a transient initiator, affecting the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment functions.
Inspection Report# : 2005003(pdf)
Significance:        Mar 24, 2005 Identified By: Self-Revealing Item Type: FIN Finding Unplanned reactor trip due to ineffective use of industry OE during a maintenance activity.
A self-revealing finding was identified after an unplanned reactor trip resulted from the licensee's ineffective use of industry operating experience. The plant tripped from low steam generator level after a feedwater regulating valve closed. The regulating valve closed after a control power supply shorted during a maintenance activity. The power supply shorted because the maintenance workers had used an inadequate work instruction. A similar event occurred at the Beaver Valley Nuclear Plant during June 2003. The licensee failed to effectively use the operating experience when planning and performing the maintenance activity. The licensee's failure to properly revise an incorrect work package before proceeding with the work activity, a poor prejob brief, and organizational time pressures also contributed to the event.
Additionally, the licensee's evaluation of the event identified contributing causes as root causes, and did not take into account the programmatic issues to include operating experience reviews into work instruction development procedures. This finding had crosscutting aspects regarding human performance, and problem identification and resolution in that the evaluation of root versus contributing causes was deficient.
This finding was more than minor because the procedural adequacy attribute of the initiating events cornerstone objective was affected. The inspectors concluded the reactor trip is a transient initiator, affecting the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment functions.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: FIN Finding Operator Error Resulted in a Steam Generator Chemistry Excursion.
A self-revealing finding was identified after an operator error resulted in an unplanned secondary side chemistry excursion and a steam generator blowdown isolation. An operator failed to maintain minimum cooling tower blowdown flow during an effluent release of steam generator blowdown demineralizer flush water to the environment. The reduction in flow resulted in the isolation of the release and pressurization of the steam generator blowdown flush line. The pressurized line resulted in the transfer of flush water to the main condenser and caused steam generator chemistry to exceed the Action Level 2 threshold. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance (personnel).
This finding is greater than minor because the chemistry excursion had an impact on the equipment performance attribute of the initiating events objective cornerstone. The inspectors determined that this finding is of very low safety significance because the chemistry excursion did not add to the likelihood of a primary or secondary system loss of coolant accident initiator, did not contribute to loss of mitigation equipment functions, and did not increase the likelihood of a fire or internal/external flood.
Inspection Report# : 2004005(pdf)
Significance: N/A Nov 08, 2004 Identified By: NRC Item Type: FIN Finding
 
3Q/2005 Inspection Findings - Callaway                                                                                                    Page 2 of 6 Supplemental Inspection for a White performance indicator in the initiating events cornerstone.
The NRC conducted a supplemental inspection to assess the licensee's evaluation of conditions associated with a White performance indicator in the initiating events cornerstone. Three unplanned reactor trips resulted in the unplanned scrams per 7,000 critical hours performance indicator to cross the threshold from Green to White during the second quarter of 2004. The inspector concluded that the licensee's problem identification, root cause, extent-of-condition evaluations, and corrective actions for the three reactor trips were adequate. Two of the reactor trips were caused by main generator supervisory relay failures. The third reactor trip was caused by a reactor operator's failure to follow the power ascension procedure. Several of the root causes contributing to the third reactor trip have been long-standing station problems. The inspector identified weaknesses in the licensee's root cause determination and corrective actions related to the third reactor trip. The inspector did not identify any common attributes linking the three reactor trips from a risk perspective.
Inspection Report# : 2004009(pdf)
Mitigating Systems Significance:        Sep 23, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Degraded auxiliary feedwater pump due to the failure to follow procedure.
A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after AmerenUE failed to properly align the turbine driven auxiliary feedwater pump mechanical overspeed trip mechanism after surveillance testing. The trip mechanism was misaligned from August 1 - 18, 2005. The misaligned trip mechanism increased the probability the turbine would trip if the pump would have been required to respond to an event. This issue was entered into the corrective action program as Callaway Action Request 200505801. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance.
This finding is greater than minor because the degraded trip mechanism affected the reactor mitigating systems cornerstone and the equipment performance attribute to ensure availability of systems that respond to prevent core damage. This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2005004(pdf)
Significance:        Jun 23, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain the integrity of a three-hour auxiliary building fire door.
A self-revealing noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee failed to maintain the integrity of an auxiliary building fire door that was required to provide a three-hour fire barrier.
This finding is greater than minor because the reactor safety mitigating systems cornerstone attribute to provide protection against external factors was affected. The inspectors used Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the degraded door is a fire barrier related to the licensee's fire protection defense-in-depth strategies. The licensee had several prior opportunities to self-identify the degraded door and previous corrective actions were not effective to prevent recurrence. The inspectors concluded that the condition was intermittent and thus had a low degradation rating. The inspectors concluded this finding is of very low safety significance because of the low degradation level.
Inspection Report# : 2005003(pdf)
Significance:        Mar 24, 2005 Identified By: NRC Item Type: NCV NonCited Violation Ineffective cause determination and corrective actions to prevent recurrence of ECCS pipe voiding.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, after the licensee's cause determination and corrective actions were ineffective to prevent recurrence of safety injection pump discharge pipe voiding. Plant Technical Specifications required the licensee to verify that the emergency core cooling system piping was full of water every 31 days. The licensee established a 20 percent maximum void fraction as the acceptance limit for the safety injection pump hot leg injection discharge piping. On seven occasions during the past 2 years the surveillance acceptance criteria was not met. This finding had crosscutting aspects regarding problem identification and resolution in that the licensee's actions to determine the cause of the repeated surveillance failures and to implement corrective actions were not effective in preventing recurrence of the condition.
This finding is greater than minor because voiding in emergency core cooling system piping affected the reactor mitigating systems cornerstone and the equipment performance attribute to ensure availability of systems that respond to prevent core damage. This finding was only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic
 
3Q/2005 Inspection Findings - Callaway                                                                                                  Page 3 of 6 Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2005002(pdf)
Significance:        Mar 24, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Unplanned loss of water fire supression due to an inadequate testing procedure.
A self revealing noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee inadvertently isolated all plant fire water suppression from the reactor, auxiliary, control, and turbine buildings during surveillance testing. The isolation resulted in the unplanned loss of all fire water to the reactor, auxiliary, control, and turbine buildings. The isolation occurred due an inadequate surveillance testing procedure. The licensee identified the isolation of the fire loops after about 15 minutes. The licensee reestablished the fire water suppression system after about 1.5 hours. This finding had crosscutting aspects regarding human performance in that the procedure used was inadequate.
The finding is greater than minor because the unplanned isolation of fire water was associated with the "Protection Against External Factors,"
attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure availability of systems designed to respond to initiating events. The inspectors used Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the condition had an adverse affect on fire defense-in-depth strategies. The senior reactor analyst evaluated the finding based on a bounding calculation for each fire area affected by the loss of fire water in the plant. The analyst concluded a plant-wide fire mitigation probability of 4.3 x 10-6 over the 2-hour exposure period. The analyst assumed that the maximum Conditional Core Damage Probability for any fire area was bounded by probability used to assess fires requiring control room evacuation (CCDP=0.1). The maximum resulting core damage probability from internal fires over the 2-hour period was the product of the plant-wide fire mitigation probability and 0.1. This bounded the risk of the finding resulting in no greater increase in core damage frequency than 4.3 x 10-7. The analyst concluded that a systematic search and assessment effort was beyond the intended scope of the fire protection significance determination process. Therefore, in accordance with NRC Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Section 05.04.c, regional management reviewed this finding and determined that it was of very low risk significance.
Inspection Report# : 2005002(pdf)
Significance:        Mar 24, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain the minimum number of fire brigade members on site.
An NRC identified noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee failed to maintain the minimum number of fire brigade members on site. The inspectors identified that the licensee did not maintain minimum fire brigade staffing. The licensee was required to maintain at least five fire brigade members on site at all times. Between January 24 and February 9, 2005, the outside equipment operator was assigned to the fire brigade 68 percent of the time. However, the outside equipment operator spent about 80 percent of the shift outside of the protected area, including attending equipment at the river pumping station, located eight miles from the site. The inspectors concluded that full fire brigade staffing would have been delayed about 20 to 30 minutes if the activation occurred while the equipment operator was performing outside duties. This finding had crosscutting aspects regarding human performance in that full fire brigade staffing was not ensured. This finding also had crosscutting aspects regarding problem identification and resolution in that the issue was not properly evaluated following documentation in the corrective action program twice.
This finding is greater than minor because the reactor safety mitigating systems cornerstone objective attribute to provide protection against external factors was affected. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," does not address fire brigade performance deficiencies. Regional management review concluded this finding was of very low safety significance because it affected the fire prevention and administrative controls category and represented only a short duration degradation in fire brigade staffing.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain the Integrity of an Auxiliary Building Fire Door The inspectors identified a noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," after the licensee failed to maintain the integrity of an auxiliary building fire door. The inspectors identified that the fire door would not provide the rated fire confinement function because of a broken latch. The door provided the 3-hour fire barrier between auxiliary building fire Areas A-19 and A-20. The licensee had several prior opportunities to identify the degraded fire door. The plant security procedure required plant security officers to verify that the fire door was properly latched during each patrol. Several security patrols passed through the door each shift. This finding had crosscutting aspects related to human performance (personnel) in that the plant procedure regarding verification of fire doors was not followed.
This finding is greater than minor because the fire door was associated with the mitigating system cornerstone attribute to provide protection against external factors. The inspectors concluded that the degraded door was a fire confinement finding with a high degradation rating due to the broken latch. This finding is of very low safety significance because the degraded door did not separate unique potential fire damage targets
 
3Q/2005 Inspection Findings - Callaway                                                                                                  Page 4 of 6 and that the door would have provided at least 20 minutes fire endurance protection. The inspectors also concluded that no fixed or in-situ fire ignition sources or combustible or flammable materials were positioned such that the degraded door would have been subject to direct flame impingement.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Required Fire Watch.
The inspectors identified a noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," after a plant fire occurred when the licensee failed to establish a required fire watch. A welder ignited a fire on the communication corridor roof. The fire burned through the roof and ignited the ceiling below. The licensee had not established a fire watch inside the room. The plant fire brigade responded and extinguished the fire. The fire brigade left the area without establishing a re-flash fire watch. About 55 minutes later, an equipment operator returned to the room and identified that the fire had reignited. The plant fire brigade responded again and extinguished the re-flash fire.
This finding is greater than minor because the mitigating systems cornerstone attribute providing protection against external factors was affected. This finding had an adverse affect on the licensee's fire protection defense-in-depth strategies related to fire detection, manual suppression, and fire brigade effectiveness. The inspectors concluded that the lack of a fire watch degraded the licensee's early fire suppression capability and resulted in the fire prevention finding with a high degradation rating. The inspectors determined that this finding is of very low significance because the fire ignition source could not have caused ignition of secondary combustible fuels and was not close enough to sufficient surrounding combustibles to cause damage consistent with any of the plant fire damage scenarios.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Postmaintenance Test Failed to Identify Degraded Turbine Driven Auxiliary Feedwater Pump Bearing Cooling Following Maintenance.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," after postmaintenance testing was not adequate to identify degraded turbine-driven auxiliary feedwater pump bearing cooling following maintenance. The licensee completed an overhaul of the turbine, performed a postmaintenance test, and returned the system to service. Twenty-four days later, the licensee observed elevated inboard turbine bearing temperatures during a surveillance test. The elevated temperatures were caused by an obstruction in the lube oil cooler. The lube oil filter had been improperly installed during the overhaul and allowed particulate material to bypass the filter. The inspectors identified that an elevated bearing temperature also occurred during the earlier postmaintenance test. However, the licensee did not monitor bearing temperatures during the test nor was postmaintenance testing performed for a sufficient duration to allow the turbine to reach normal operating temperatures. This finding had crosscutting aspects regarding human performance (personnel) for failure to adequately test the turbine-driven auxiliary feedwater pump following maintenance, and problem identification in that indications were present during an earlier test that should have alerted the licensee to the condition.
This finding is greater than minor because, if left uncorrected the condition would become a more significant safety concern. This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2004005(pdf)
Barrier Integrity Significance: TBD Sep 23, 2005 Identified By: Self-Revealing Item Type: AV Apparent Violation Operator error resulted in the loss of configuration control while shutdown.
A self-revealing apparent violation of Technical Specification 5.4.1.a, "Procedures," was identified after an operator error resulted in the failure to maintain the required cold overpressure mitigation system configuration while the reactor was in Mode 5. Technical Specification 3.4.12, "Cold Overpressure Mitigation System," prohibited more than one centrifugal charging pump from being capable of injecting into the reactor vessel. An operator inadvertently defeated administrative controls and enabled a centrifugal charging pump during a diesel generator and sequencer test restoration lineup on September 20, 2005. Contributing causes to the event were inadequate procedural controls and pre-job brief. This issue was entered into the corrective action program as Callaway Action Request 200507092. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance.
This finding is greater than minor because, if left uncorrected, it would have become a more significant safety concern involving the integrity of the reactor coolant system boundary (barrier integrity cornerstone). The finding was evaluated using Manual Chapter 0609, "Significance
 
3Q/2005 Inspection Findings - Callaway                                                                                                Page 5 of 6 Determination Process," Appendix G, Shutdown Operations Significance, Checklist 2. Although the performance deficiency did not result in a Technical Specification violation, discussions with the Office of Nuclear Reactor Regulation identified a Phase 3 analysis should be performed and is currently under evaluation.
Inspection Report# : 2005004(pdf)
Significance:        Sep 23, 2005 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions resulted in degraded control building habitability boundary.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, "Corrective Action," after ineffective corrective actions resulted in a repeat degradation of a control building emergency ventilation habitability boundary door. AmerenUE's work control organization twice authorized work on the essential switchgear room to emergency diesel generator room door without approval of the shift operations department. As a result, shift operations did not understand that the habitability boundary had been compromised by the maintenance. This finding, which involved ineffective corrective actions to prevent the repeat degradation of the ventilation system habitability boundary door, was associated with the crosscutting area of problem identification and resolution.
This finding was greater than minor because it was associated with the integrity of the control building pressure envelope in that the degraded door would not meet its habitability function. The finding was only of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2005004(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation An Operator Error Resulted in an Unplanned Transfer of Water from Spent Fuel Pool.
A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an operator error resulted in the unplanned transfer of 3600 gallons of water from the spent fuel pool. The operating procedure required the operator to shutdown refueling water storage tank recirculation before placing fuel pool cleanup in service. The operator failed to shutdown the recirculation lineup resulting in the unplanned spent fuel pool water loss. The operating crew recognized the decreasing spent fuel pool level and secured the recirculation after about 3600 gallons had been transferred.
This finding is greater than minor because if left uncorrected it would have become a more significant safety concern. The inspectors determined that this finding is only of very low significance because the condition only represented a degradation of the radiological barrier function provided by the spent fuel pool.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Feedwater Isolation Valve Post Modification Deficiencies.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," after the licensee failed to correct deficiencies identified during post modification testing of the feedwater isolation valve actuators. The post modification test revealed that the feedwater isolation valves would not meet the Mode 3 closure times described in the licensing bases. The licensee dispositioned the deficiency without adequately correcting the deficiencies. The licensee had a second opportunity to identify the inadequate corrective actions when the Independent Technical Review Team assessed the post modification test results. The Independent Technical Review Team assessment was not effective in identifying the inadequate corrective actions. This finding has crosscutting aspects regarding failure to implement adequate corrective actions.
This finding is greater than minor because the failure of the feedwater isolation valves to meet closures times affected the barrier integrity cornerstone design control attribute to maintain the functionality of the fuel cladding, following a cooldown event, and to limit post accident containment pressure by isolating feedwater to the faulted steam generator. This finding is only of very low safety significance because the condition did not represent a degradation of the barrier function of the control room, auxiliary building, or spent fuel pool, nor did this finding represent an actual open pathway in the physical integrity of the containment, nor affect the atmospheric pressure control or hydrogen control functions of containment.
Inspection Report# : 2004005(pdf)
Emergency Preparedness
 
3Q/2005 Inspection Findings - Callaway                                                                                              Page 6 of 6 Occupational Radiation Safety Significance:        Jun 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 20.1201(f) for failure to reduce individuals exposure margin.
The team identified a non-cited violation of 10 CFR 20.1201(f) when the licensee failed to reduce the dose that individuals may be allowed to receive in the current year by the amount of occupational dose received at other facilities. Specifically, on May 16, 2005, the licensee failed to enter inspectors' year-to-date exposure into the PRORAD computer system and subsequently reduce their allowable exposure margin.
The finding is greater than minor because it was associated with a Occupational Radiation Safety cornerstone attribute (Program & Process) and it affected the associated cornerstone objective. The failure to reduce exposure margins to control personnel exposure decreases the licensee's ability to ensure adequate protection of the worker health and safety from exposure to radiation. The significance of the finding was evaluated using the Occupational Radiation Safety Significance Determination Process because the finding involved an individual worker's potential for unplanned, unintended dose resulting from actions contrary to NRC regulations. The finding was determined to be of very low safety significance because the finding did not involve; (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for an overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had cross-cutting aspects associated with human performance. Licensee personnel directly contributed to the finding when they failed to enter workers' exposure into the licensee's dose tracking computer system. The finding was placed into the licensee's corrective action program as CAR 2005-03354.
Inspection Report# : 2005011(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : November 30, 2005
 
4Q/2005 Inspection Findings - Callaway                                                                                                  Page 1 of 6 Callaway 4Q/2005 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Unplanned auxiliary feedwater actuation due to use of an inadequate general operating procedure for troubleshooting.
A self-revealing noncited violation of Technical Specification 5.4.1.a was identified after an unplanned auxiliary feedwater actuation and reactor trip signal occurred while shutdown due to an inadequate general operating procedure and poor crew decision making.
This finding is greater than minor because the procedural adequacy attribute of the initiating events cornerstone objective is affected. The inspectors concluded the auxiliary feedwater actuation and reactor trip signal was a transient initiator, affecting the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment functions.
Inspection Report# : 2005003(pdf)
Significance:        Mar 24, 2005 Identified By: Self-Revealing Item Type: FIN Finding Unplanned reactor trip due to ineffective use of industry OE during a maintenance activity.
A self-revealing finding was identified after an unplanned reactor trip resulted from the licensee's ineffective use of industry operating experience. The plant tripped from low steam generator level after a feedwater regulating valve closed. The regulating valve closed after a control power supply shorted during a maintenance activity. The power supply shorted because the maintenance workers had used an inadequate work instruction. A similar event occurred at the Beaver Valley Nuclear Plant during June 2003. The licensee failed to effectively use the operating experience when planning and performing the maintenance activity. The licensee's failure to properly revise an incorrect work package before proceeding with the work activity, a poor prejob brief, and organizational time pressures also contributed to the event.
Additionally, the licensee's evaluation of the event identified contributing causes as root causes, and did not take into account the programmatic issues to include operating experience reviews into work instruction development procedures. This finding had crosscutting aspects regarding human performance, and problem identification and resolution in that the evaluation of root versus contributing causes was deficient.
This finding was more than minor because the procedural adequacy attribute of the initiating events cornerstone objective was affected. The inspectors concluded the reactor trip is a transient initiator, affecting the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment functions.
Inspection Report# : 2005002(pdf)
Mitigating Systems Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Minimum gap size exceeded for containment recirculation sump.
The inspectors identified a noncited violation of 10 CFR Part 50, Criterion X, after plant quality control personnel performed an inadequate inspection of an emergency core cooling system containment recirculation sump. The inspection failed to identify a 11/2-inch hole which provided a path for foreign material into the containment sump which could affect the recirculation mode of emergency core cooling system operation. AmerenUE completed a detailed inspection of the sump on April 27, 2004 in response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors," but failed to identify the 11/2 -inch hole. This issue was entered into the corrective action program as Callaway Action Request 200509189.
This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure availability and reliability of the containment recirculation sump emergency core cooling system containment safety function. This finding is of very low safety significance because the condition was a qualification deficiency confirmed not to result in loss of function per Part 9900, Technical Assessment; "Operability Determination Process for Operability and
 
4Q/2005 Inspection Findings - Callaway                                                                                                  Page 2 of 6 Functional Assessment." The cause of this finding, poor attention to detail by personnel, is related to the crosscutting element of human performance.
Inspection Report# : 2005005(pdf)
Significance:          Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately implement continuous compensatory fire watches.
The inspectors identified a noncited violation of Technical Specification 5.4.1, "Procedures," associated with seven examples of inadequately performed continuous fire watches. In September 2005, AmerenUE provided verbal guidance to fire watch personnel that continuous watches may be met by a 15 minute roving fire patrol. The roving patrol did not ensure adequate compensatory action for fire areas with degraded detection or suppression capability. As a result, fire watch personnel were not available to promptly detect, report, and extinguish a fire while still in the incipient stage. AmerenUE did not evaluate this change to ensure no adverse affect on the ability to achieve and maintain safe shutdown in the event a fire was created. The condition was entered into the corrective action program as Callaway Action Request 200510325.
The cause of this finding is related to the crosscutting element of human performance because the resources needed to support the task, including complete and accurate procedures and supervision, were less than adequate.
This finding is greater than minor because inadequate fire watches are associated with the reactor safety mitigating systems cornerstone attribute to provide protection against external factors and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding is of very low safety significance because the condition had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for outages or degradations. A low degradation rating was assigned to this finding as the provision affected by this finding is expected to display nearly the same level of effectiveness and reliability.
Inspection Report# : 2005005(pdf)
Significance:          Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Conduct Simulator Testing in Accordance with ANSI/ANS 3.5-1998 The inspectors determined that the failure to adhere to ANSI/ANS 3.5-1998, as endorsed by Regulatory Guide 1.149 "Nuclear Power Plant Simulation Facilities for Use in Operator Training and License Examinations,"Revision 3, October 2001, as committed to in the Callaway Plant Simulation certification dated March 13, 2000, was a finding. Specifically, the simulator performance testing did not meet the standards specified in ANSI/ANS 3.5-1998, in that: (1) all required parameters during the simulator test were not recorded; and (2) simulator to baseline data comparisons were unavailable.
The failure to evaluate and document simulator performance testing is more than minor because it affected the Operator Requalification attribute of the Mitigating Systems and Initiating Event cornerstone of reactor safety and is inconsistent with the requirements of 10 CFR 55.46 in that simulator fidelity issues may not be identified, which have the potential of causing negative training. The finding was considered to be of very low safety significance because the discrepancies have not yet impacted operator actions in the plant, such that, safety-related equipment was made inoperable or that operators failed to properly respond to plant transients.
Inspection Report# : 2005005(pdf)
Significance:          Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Use of a Non-Qualified Calculation in a Safety Related Modification The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," associated with an inadequate engineering procedure used for the verification of design calculations. The inadequate procedure resulted in a non-qualified, non-safety-related engineering calculation being used to demonstrate that the safety-related containment recirculation sump valves were capable of performing the safety function described in the design bases. The performance deficiency associated with this finding involved the failure of engineering personnel to only use qualified calculations for safety-related applications. The cause of this finding is related to the crosscutting element of human performance because insufficient resources were provided to ensure complete and accurate procedures to support task performance. This finding was entered into the Corrective Action Program as Callaway Action Request 200509849.
This finding is greater than minor because if left uncorrected, this finding would become a more significant safety concern. This finding is determined to have very low safety significance because this issue involves a design deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessment."
Inspection Report# : 2005005(pdf)
 
4Q/2005 Inspection Findings - Callaway                                                                                                    Page 3 of 6 Significance:        Sep 23, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Degraded auxiliary feedwater pump due to the failure to follow procedure.
A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after AmerenUE failed to properly align the turbine driven auxiliary feedwater pump mechanical overspeed trip mechanism after surveillance testing. The trip mechanism was misaligned from August 1 - 18, 2005. The misaligned trip mechanism increased the probability the turbine would trip if the pump would have been required to respond to an event. This issue was entered into the corrective action program as Callaway Action Request 200505801. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance.
This finding is greater than minor because the degraded trip mechanism affected the reactor mitigating systems cornerstone and the equipment performance attribute to ensure availability of systems that respond to prevent core damage. This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2005004(pdf)
Significance:        Jun 23, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain the integrity of a three-hour auxiliary building fire door.
A self-revealing noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee failed to maintain the integrity of an auxiliary building fire door that was required to provide a three-hour fire barrier.
This finding is greater than minor because the reactor safety mitigating systems cornerstone attribute to provide protection against external factors was affected. The inspectors used Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the degraded door is a fire barrier related to the licensee's fire protection defense-in-depth strategies. The licensee had several prior opportunities to self-identify the degraded door and previous corrective actions were not effective to prevent recurrence. The inspectors concluded that the condition was intermittent and thus had a low degradation rating. The inspectors concluded this finding is of very low safety significance because of the low degradation level.
Inspection Report# : 2005003(pdf)
Significance:        Mar 24, 2005 Identified By: NRC Item Type: NCV NonCited Violation Ineffective cause determination and corrective actions to prevent recurrence of ECCS pipe voiding.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, after the licensee's cause determination and corrective actions were ineffective to prevent recurrence of safety injection pump discharge pipe voiding. Plant Technical Specifications required the licensee to verify that the emergency core cooling system piping was full of water every 31 days. The licensee established a 20 percent maximum void fraction as the acceptance limit for the safety injection pump hot leg injection discharge piping. On seven occasions during the past 2 years the surveillance acceptance criteria was not met. This finding had crosscutting aspects regarding problem identification and resolution in that the licensee's actions to determine the cause of the repeated surveillance failures and to implement corrective actions were not effective in preventing recurrence of the condition.
This finding is greater than minor because voiding in emergency core cooling system piping affected the reactor mitigating systems cornerstone and the equipment performance attribute to ensure availability of systems that respond to prevent core damage. This finding was only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2005002(pdf)
Significance:        Mar 24, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Unplanned loss of water fire supression due to an inadequate testing procedure.
A self revealing noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee inadvertently isolated all plant fire water suppression from the reactor, auxiliary, control, and turbine buildings during surveillance testing. The isolation resulted in the unplanned loss of all fire water to the reactor, auxiliary, control, and turbine buildings. The isolation occurred due an inadequate surveillance testing procedure. The licensee identified the isolation of the fire loops after about 15 minutes. The licensee reestablished the fire water suppression system after about 1.5 hours. This finding had crosscutting aspects regarding human performance in that the procedure used was inadequate.
 
4Q/2005 Inspection Findings - Callaway                                                                                                Page 4 of 6 The finding is greater than minor because the unplanned isolation of fire water was associated with the "Protection Against External Factors,"
attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure availability of systems designed to respond to initiating events. The inspectors used Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the condition had an adverse affect on fire defense-in-depth strategies. The senior reactor analyst evaluated the finding based on a bounding calculation for each fire area affected by the loss of fire water in the plant. The analyst concluded a plant-wide fire mitigation probability of 4.3 x 10-6 over the 2-hour exposure period. The analyst assumed that the maximum Conditional Core Damage Probability for any fire area was bounded by probability used to assess fires requiring control room evacuation (CCDP=0.1). The maximum resulting core damage probability from internal fires over the 2-hour period was the product of the plant-wide fire mitigation probability and 0.1. This bounded the risk of the finding resulting in no greater increase in core damage frequency than 4.3 x 10-7. The analyst concluded that a systematic search and assessment effort was beyond the intended scope of the fire protection significance determination process. Therefore, in accordance with NRC Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Section 05.04.c, regional management reviewed this finding and determined that it was of very low risk significance.
Inspection Report# : 2005002(pdf)
Significance:        Mar 24, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain the minimum number of fire brigade members on site.
An NRC identified noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee failed to maintain the minimum number of fire brigade members on site. The inspectors identified that the licensee did not maintain minimum fire brigade staffing. The licensee was required to maintain at least five fire brigade members on site at all times. Between January 24 and February 9, 2005, the outside equipment operator was assigned to the fire brigade 68 percent of the time. However, the outside equipment operator spent about 80 percent of the shift outside of the protected area, including attending equipment at the river pumping station, located eight miles from the site. The inspectors concluded that full fire brigade staffing would have been delayed about 20 to 30 minutes if the activation occurred while the equipment operator was performing outside duties. This finding had crosscutting aspects regarding human performance in that full fire brigade staffing was not ensured. This finding also had crosscutting aspects regarding problem identification and resolution in that the issue was not properly evaluated following documentation in the corrective action program twice.
This finding is greater than minor because the reactor safety mitigating systems cornerstone objective attribute to provide protection against external factors was affected. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," does not address fire brigade performance deficiencies. Regional management review concluded this finding was of very low safety significance because it affected the fire prevention and administrative controls category and represented only a short duration degradation in fire brigade staffing.
Inspection Report# : 2005002(pdf)
Barrier Integrity Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulted in Violation of RCS Cooldown and Heatup Rate Limits.
The inspectors identified a noncited violation of Technical Specification 5.4.1.a, "Procedures," after AmerenUE Operations personnel failed to maintain the reactor coolant system temperature limits on two occasions. On November 7, 2005, plant operators decreased the reactor coolant system pressurizer surge line temperature 260 degrees Fahrenheit in a one-hour period. The operators conducted the rapid cooldown after several containment lead shield blanket polyvinylchloride covers left in containment melted. On November 8, 2005, plant operators increased the surge line temperature about 175 degrees Fahrenheit in a one-hour period. Plant Technical Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and Plant procedures required reactor coolant system component temperature changes (except the pressurizer) be limited to 100 degrees in one hour. The cause of this finding is related to the crosscutting element of human performance because of personnel failure to follow procedures.
This finding was greater than minor because it is associated with the reactor safety barrier integrity cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure reasonable assurance that the reactor coolant system piping barrier will protect the public from radionuclide releases caused by accidents or events. This finding is determined to have very low safety significance because an engineering evaluation concluded that the temperature transient did not significantly increase the likelihood of a loss of reactor coolant system inventory or degrade the ability to terminate a leak path. This finding was placed in the Corrective Action Program as Callaway Action Requests 200509487 and 200509143.
Inspection Report# : 2005005(pdf)
Significance:        Sep 23, 2005 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2005 Inspection Findings - Callaway                                                                                                Page 5 of 6 Ineffective corrective actions resulted in degraded control building habitability boundary.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, "Corrective Action," after ineffective corrective actions resulted in a repeat degradation of a control building emergency ventilation habitability boundary door. AmerenUE's work control organization twice authorized work on the essential switchgear room to emergency diesel generator room door without approval of the shift operations department. As a result, shift operations did not understand that the habitability boundary had been compromised by the maintenance. This finding, which involved ineffective corrective actions to prevent the repeat degradation of the ventilation system habitability boundary door, was associated with the crosscutting area of problem identification and resolution.
This finding was greater than minor because it was associated with the integrity of the control building pressure envelope in that the degraded door would not meet its habitability function. The finding was only of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2005004(pdf)
Emergency Preparedness Significance:        Jan 12, 2005 Identified By: NRC Item Type: NCV NonCited Violation Change in Emergency Action Level 3E decreased the effectiveness of the Emergency Plan The inspector identified a violation of 10 CFR 50.54(q) for implementing a change to emergency action levels, which decreased the effectiveness of the emergency plan. Emergency Implementing Plan Procedure EIP-ZZ-00101, "Classifying the Emergency," Revision 33, limited application of emergency action Level 3E, "Fire within Protected Area Boundary NOT Extinguished with 15 minutes of Verification" so that fires in some plant areas which would be classified under the previous revision may no longer be classifiable.
Implementation of changes to emergency action levels, which decreased the effectiveness of the emergency plan was a performance deficiency.
The finding is more than minor because removal of a classifiable condition from licensee emergency action levels has the potential to impact safety, and licensee implementation of a change to their emergency plan, which decreases the effectiveness of the plan without prior NRC approval, impacts the regulatory process. This finding is a violation of 10 CFR 50.54(q). The licensee has entered this issue into their corrective action system as Corrective Action Report 200510162.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Significance:        Oct 21, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to control a high radiation area with dose rates greater than 1.0 rem per hour.
The inspector reviewed a self-revealing non-cited violation of Technical Specification 5.7.2 because the licensee failed to control a high radiation area with dose rates greater than 1.0 rem per hour. Specifically, on September 26, 2005, the reactor vessel head was moved from the head stand and placed back on the reactor vessel without the proper radiological controls in place for a high radiation area with dose rates as high as 6.0 rem per hour. A loud noise created by the falling of a locking device on the reactor head alerted radiation protection personnel that the head lift had begun prematurely. The licensee's immediate corrective actions were to ensure that individuals were not present in the high radiation area and to place the reactor head in a safe condition on the reactor vessel. The finding was entered into the licensee's corrective action program as Callaway Action Request 200507546.
The failure to control a high radiation area with dose rates greater than 1.0 rem per hour is a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involved the potential for a worker's unplanned or unintended dose resulting from actions contrary to technical specifications. When processed through the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because the finding did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. In addition, this finding has crosscutting aspects associated with human performance because poor coordination and communication between the head lift crew and radiation protection personnel directly contributed to the finding.
Inspection Report# : 2005009(pdf)
Significance:        Jun 02, 2005
 
4Q/2005 Inspection Findings - Callaway                                                                                              Page 6 of 6 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 20.1201(f) for failure to reduce individuals exposure margin.
The team identified a non-cited violation of 10 CFR 20.1201(f) when the licensee failed to reduce the dose that individuals may be allowed to receive in the current year by the amount of occupational dose received at other facilities. Specifically, on May 16, 2005, the licensee failed to enter inspectors' year-to-date exposure into the PRORAD computer system and subsequently reduce their allowable exposure margin.
The finding is greater than minor because it was associated with a Occupational Radiation Safety cornerstone attribute (Program & Process) and it affected the associated cornerstone objective. The failure to reduce exposure margins to control personnel exposure decreases the licensee's ability to ensure adequate protection of the worker health and safety from exposure to radiation. The significance of the finding was evaluated using the Occupational Radiation Safety Significance Determination Process because the finding involved an individual worker's potential for unplanned, unintended dose resulting from actions contrary to NRC regulations. The finding was determined to be of very low safety significance because the finding did not involve; (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for an overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had cross-cutting aspects associated with human performance. Licensee personnel directly contributed to the finding when they failed to enter workers' exposure into the licensee's dose tracking computer system. The finding was placed into the licensee's corrective action program as CAR 2005-03354.
Inspection Report# : 2005011(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:          Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Less Than Adequate Spent Fuel Pool Water Inventory Risk Controls The inspectors identified a finding after AmerenUE implemented less than adequate risk management controls of the spent fuel pool water inventory. On September 29, 2005, the core had been off-loaded to the spent fuel pool and the transfer canal weir was removed. The spent fuel pool temperature was 99 degrees Fahrenheit with a 12.1 hour time-to-boil. Transfer tube Valve ECV-995 isolated the fuel transfer canal from the containment cavity. In this configuration, the tube valve could provide a drain path reducing water level from 25 feet to less than 2 feet above the spent fuel. Valve ECV-995 was closed but was not identified in the shutdown risk management system and did not have administrative controls to protect against misalignment. NRC Information Notice 2005-16, "Outage Planning and Scheduling - Impacts on Risk," emphasized that most spent fuel pool events had a common thread of human error and involved equipment misalignment. This finding was entered into the Corrective Action Program as Callaway Action Requests 200507593 and 200507693.
This finding is greater than minor because if left uncorrected, it would have become a more significant safety concern. Because Manual Chapter 0609, "Significance Determination Process," does not specifically address findings related to the spent fuel pool inventory, this finding is determined to have very low safety significance based on NRC management review with input from senior reactor analysts. No violation of regulatory requirements occurred.
Inspection Report# : 2005005(pdf)
Last modified : March 03, 2006
 
1Q/2006 Inspection Findings - Callaway                                                                                                  Page 1 of 5 Callaway 1Q/2006 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Unplanned auxiliary feedwater actuation due to use of an inadequate general operating procedure for troubleshooting.
A self-revealing noncited violation of Technical Specification 5.4.1.a was identified after an unplanned auxiliary feedwater actuation and reactor trip signal occurred while shutdown due to an inadequate general operating procedure and poor crew decision making.
This finding is greater than minor because the procedural adequacy attribute of the initiating events cornerstone objective is affected. The inspectors concluded the auxiliary feedwater actuation and reactor trip signal was a transient initiator, affecting the initiating events cornerstone. The inspectors determined this finding to be of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment functions.
Inspection Report# : 2005003(pdf)
Mitigating Systems Significance:        Mar 24, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of Degraded Plant Equipment The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," after the licensee failed to promptly identify, evaluate, and correct a degraded control building air conditioning unit compressor. The compressor developed a hole in one of the cylinder head discharge reed valves. The hole allowed the bypass of hot discharge gases and rendered the compressor incapable of completing the safety function for the specified mission time. The hole was caused by cyclic fatigue stress. This issue was entered into the corrective action program as Callaway Action Request 200601177. This finding is associated with the crosscutting area of problem identification and resolution because the issue involved the failure of operations personnel to adequately evaluate degraded plant equipment.
This finding is greater than minor because, if left uncorrected, the degradation would have worsened and become a more significant safety concern. This finding was a qualification deficiency that resulted in loss of operability per "Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment." However, the finding is of very low safety significance because it did not represent a loss of system safety function, did not represent an actual loss of safety function for a single train for greater than the 30-day Technical Specification allowed outage time, did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk-significant per 10 CFR 50.65, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Minimum gap size exceeded for containment recirculation sump.
The inspectors identified a noncited violation of 10 CFR Part 50, Criterion X, after plant quality control personnel performed an inadequate inspection of an emergency core cooling system containment recirculation sump. The inspection failed to identify a 11/2-inch hole which provided a path for foreign material into the containment sump which could affect the recirculation mode of emergency core cooling system operation. AmerenUE completed a detailed inspection of the sump on April 27, 2004 in response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors," but failed to identify the 11/2 -inch hole. This issue was entered into the corrective action program as Callaway Action Request 200509189.
This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure availability and reliability of the containment recirculation sump emergency core cooling system containment safety function. This finding is of very low safety significance because the condition was a qualification deficiency confirmed not to result in loss of function per Part 9900, Technical Assessment; "Operability Determination Process for Operability and
 
1Q/2006 Inspection Findings - Callaway                                                                                                  Page 2 of 5 Functional Assessment." The cause of this finding, poor attention to detail by personnel, is related to the crosscutting element of human performance.
Inspection Report# : 2005005(pdf)
Significance:          Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately implement continuous compensatory fire watches.
The inspectors identified a noncited violation of Technical Specification 5.4.1, "Procedures," associated with seven examples of inadequately performed continuous fire watches. In September 2005, AmerenUE provided verbal guidance to fire watch personnel that continuous watches may be met by a 15 minute roving fire patrol. The roving patrol did not ensure adequate compensatory action for fire areas with degraded detection or suppression capability. As a result, fire watch personnel were not available to promptly detect, report, and extinguish a fire while still in the incipient stage. AmerenUE did not evaluate this change to ensure no adverse affect on the ability to achieve and maintain safe shutdown in the event a fire was created. The condition was entered into the corrective action program as Callaway Action Request 200510325.
The cause of this finding is related to the crosscutting element of human performance because the resources needed to support the task, including complete and accurate procedures and supervision, were less than adequate.
This finding is greater than minor because inadequate fire watches are associated with the reactor safety mitigating systems cornerstone attribute to provide protection against external factors and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding is of very low safety significance because the condition had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for outages or degradations. A low degradation rating was assigned to this finding as the provision affected by this finding is expected to display nearly the same level of effectiveness and reliability.
Inspection Report# : 2005005(pdf)
Significance:          Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Conduct Simulator Testing in Accordance with ANSI/ANS 3.5-1998 The inspectors determined that the failure to adhere to ANSI/ANS 3.5-1998, as endorsed by Regulatory Guide 1.149 "Nuclear Power Plant Simulation Facilities for Use in Operator Training and License Examinations,"Revision 3, October 2001, as committed to in the Callaway Plant Simulation certification dated March 13, 2000, was a finding. Specifically, the simulator performance testing did not meet the standards specified in ANSI/ANS 3.5-1998, in that: (1) all required parameters during the simulator test were not recorded; and (2) simulator to baseline data comparisons were unavailable.
The failure to evaluate and document simulator performance testing is more than minor because it affected the Operator Requalification attribute of the Mitigating Systems and Initiating Event cornerstone of reactor safety and is inconsistent with the requirements of 10 CFR 55.46 in that simulator fidelity issues may not be identified, which have the potential of causing negative training. The finding was considered to be of very low safety significance because the discrepancies have not yet impacted operator actions in the plant, such that, safety-related equipment was made inoperable or that operators failed to properly respond to plant transients.
Inspection Report# : 2005005(pdf)
Significance:          Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Use of a Non-Qualified Calculation in a Safety Related Modification The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," associated with an inadequate engineering procedure used for the verification of design calculations. The inadequate procedure resulted in a non-qualified, non-safety-related engineering calculation being used to demonstrate that the safety-related containment recirculation sump valves were capable of performing the safety function described in the design bases. The performance deficiency associated with this finding involved the failure of engineering personnel to only use qualified calculations for safety-related applications. The cause of this finding is related to the crosscutting element of human performance because insufficient resources were provided to ensure complete and accurate procedures to support task performance. This finding was entered into the Corrective Action Program as Callaway Action Request 200509849.
This finding is greater than minor because if left uncorrected, this finding would become a more significant safety concern. This finding is determined to have very low safety significance because this issue involves a design deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessment."
Inspection Report# : 2005005(pdf)
 
1Q/2006 Inspection Findings - Callaway                                                                                                    Page 3 of 5 Significance:        Sep 23, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Degraded auxiliary feedwater pump due to the failure to follow procedure.
A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after AmerenUE failed to properly align the turbine driven auxiliary feedwater pump mechanical overspeed trip mechanism after surveillance testing. The trip mechanism was misaligned from August 1 - 18, 2005. The misaligned trip mechanism increased the probability the turbine would trip if the pump would have been required to respond to an event. This issue was entered into the corrective action program as Callaway Action Request 200505801. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance.
This finding is greater than minor because the degraded trip mechanism affected the reactor mitigating systems cornerstone and the equipment performance attribute to ensure availability of systems that respond to prevent core damage. This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2005004(pdf)
Significance:        Jun 23, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain the integrity of a three-hour auxiliary building fire door.
A self-revealing noncited violation of Technical Specification 5.4.1.d, "Fire Protection Program," was identified after the licensee failed to maintain the integrity of an auxiliary building fire door that was required to provide a three-hour fire barrier.
This finding is greater than minor because the reactor safety mitigating systems cornerstone attribute to provide protection against external factors was affected. The inspectors used Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," to analyze this finding because the degraded door is a fire barrier related to the licensee's fire protection defense-in-depth strategies. The licensee had several prior opportunities to self-identify the degraded door and previous corrective actions were not effective to prevent recurrence. The inspectors concluded that the condition was intermittent and thus had a low degradation rating. The inspectors concluded this finding is of very low safety significance because of the low degradation level.
Inspection Report# : 2005003(pdf)
Barrier Integrity Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulted in Violation of RCS Cooldown and Heatup Rate Limits.
The inspectors identified a noncited violation of Technical Specification 5.4.1.a, "Procedures," after AmerenUE Operations personnel failed to maintain the reactor coolant system temperature limits on two occasions. On November 7, 2005, plant operators decreased the reactor coolant system pressurizer surge line temperature 260 degrees Fahrenheit in a one-hour period. The operators conducted the rapid cooldown after several containment lead shield blanket polyvinylchloride covers left in containment melted. On November 8, 2005, plant operators increased the surge line temperature about 175 degrees Fahrenheit in a one-hour period. Plant Technical Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and Plant procedures required reactor coolant system component temperature changes (except the pressurizer) be limited to 100 degrees in one hour. The cause of this finding is related to the crosscutting element of human performance because of personnel failure to follow procedures.
This finding was greater than minor because it is associated with the reactor safety barrier integrity cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure reasonable assurance that the reactor coolant system piping barrier will protect the public from radionuclide releases caused by accidents or events. This finding is determined to have very low safety significance because an engineering evaluation concluded that the temperature transient did not significantly increase the likelihood of a loss of reactor coolant system inventory or degrade the ability to terminate a leak path. This finding was placed in the Corrective Action Program as Callaway Action Requests 200509487 and 200509143.
Inspection Report# : 2005005(pdf)
Significance:        Sep 23, 2005 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2006 Inspection Findings - Callaway                                                                                                Page 4 of 5 Ineffective corrective actions resulted in degraded control building habitability boundary.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, "Corrective Action," after ineffective corrective actions resulted in a repeat degradation of a control building emergency ventilation habitability boundary door. AmerenUE's work control organization twice authorized work on the essential switchgear room to emergency diesel generator room door without approval of the shift operations department. As a result, shift operations did not understand that the habitability boundary had been compromised by the maintenance. This finding, which involved ineffective corrective actions to prevent the repeat degradation of the ventilation system habitability boundary door, was associated with the crosscutting area of problem identification and resolution.
This finding was greater than minor because it was associated with the integrity of the control building pressure envelope in that the degraded door would not meet its habitability function. The finding was only of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2005004(pdf)
Emergency Preparedness Significance:        Jan 12, 2005 Identified By: NRC Item Type: NCV NonCited Violation Change in Emergency Action Level 3E decreased the effectiveness of the Emergency Plan The inspector identified a violation of 10 CFR 50.54(q) for implementing a change to emergency action levels, which decreased the effectiveness of the emergency plan. Emergency Implementing Plan Procedure EIP-ZZ-00101, "Classifying the Emergency," Revision 33, limited application of emergency action Level 3E, "Fire within Protected Area Boundary NOT Extinguished with 15 minutes of Verification" so that fires in some plant areas which would be classified under the previous revision may no longer be classifiable.
Implementation of changes to emergency action levels, which decreased the effectiveness of the emergency plan was a performance deficiency.
The finding is more than minor because removal of a classifiable condition from licensee emergency action levels has the potential to impact safety, and licensee implementation of a change to their emergency plan, which decreases the effectiveness of the plan without prior NRC approval, impacts the regulatory process. This finding is a violation of 10 CFR 50.54(q). The licensee has entered this issue into their corrective action system as Corrective Action Report 200510162.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Significance:        Oct 21, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to control a high radiation area with dose rates greater than 1.0 rem per hour.
The inspector reviewed a self-revealing non-cited violation of Technical Specification 5.7.2 because the licensee failed to control a high radiation area with dose rates greater than 1.0 rem per hour. Specifically, on September 26, 2005, the reactor vessel head was moved from the head stand and placed back on the reactor vessel without the proper radiological controls in place for a high radiation area with dose rates as high as 6.0 rem per hour. A loud noise created by the falling of a locking device on the reactor head alerted radiation protection personnel that the head lift had begun prematurely. The licensee's immediate corrective actions were to ensure that individuals were not present in the high radiation area and to place the reactor head in a safe condition on the reactor vessel. The finding was entered into the licensee's corrective action program as Callaway Action Request 200507546.
The failure to control a high radiation area with dose rates greater than 1.0 rem per hour is a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involved the potential for a worker's unplanned or unintended dose resulting from actions contrary to technical specifications. When processed through the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because the finding did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. In addition, this finding has crosscutting aspects associated with human performance because poor coordination and communication between the head lift crew and radiation protection personnel directly contributed to the finding.
Inspection Report# : 2005009(pdf)
Significance:        Jun 02, 2005
 
1Q/2006 Inspection Findings - Callaway                                                                                              Page 5 of 5 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 20.1201(f) for failure to reduce individuals exposure margin.
The team identified a non-cited violation of 10 CFR 20.1201(f) when the licensee failed to reduce the dose that individuals may be allowed to receive in the current year by the amount of occupational dose received at other facilities. Specifically, on May 16, 2005, the licensee failed to enter inspectors' year-to-date exposure into the PRORAD computer system and subsequently reduce their allowable exposure margin.
The finding is greater than minor because it was associated with a Occupational Radiation Safety cornerstone attribute (Program & Process) and it affected the associated cornerstone objective. The failure to reduce exposure margins to control personnel exposure decreases the licensee's ability to ensure adequate protection of the worker health and safety from exposure to radiation. The significance of the finding was evaluated using the Occupational Radiation Safety Significance Determination Process because the finding involved an individual worker's potential for unplanned, unintended dose resulting from actions contrary to NRC regulations. The finding was determined to be of very low safety significance because the finding did not involve; (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for an overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had cross-cutting aspects associated with human performance. Licensee personnel directly contributed to the finding when they failed to enter workers' exposure into the licensee's dose tracking computer system. The finding was placed into the licensee's corrective action program as CAR 2005-03354.
Inspection Report# : 2005011(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:          Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Less Than Adequate Spent Fuel Pool Water Inventory Risk Controls The inspectors identified a finding after AmerenUE implemented less than adequate risk management controls of the spent fuel pool water inventory. On September 29, 2005, the core had been off-loaded to the spent fuel pool and the transfer canal weir was removed. The spent fuel pool temperature was 99 degrees Fahrenheit with a 12.1 hour time-to-boil. Transfer tube Valve ECV-995 isolated the fuel transfer canal from the containment cavity. In this configuration, the tube valve could provide a drain path reducing water level from 25 feet to less than 2 feet above the spent fuel. Valve ECV-995 was closed but was not identified in the shutdown risk management system and did not have administrative controls to protect against misalignment. NRC Information Notice 2005-16, "Outage Planning and Scheduling - Impacts on Risk," emphasized that most spent fuel pool events had a common thread of human error and involved equipment misalignment. This finding was entered into the Corrective Action Program as Callaway Action Requests 200507593 and 200507693.
This finding is greater than minor because if left uncorrected, it would have become a more significant safety concern. Because Manual Chapter 0609, "Significance Determination Process," does not specifically address findings related to the spent fuel pool inventory, this finding is determined to have very low safety significance based on NRC management review with input from senior reactor analysts. No violation of regulatory requirements occurred.
Inspection Report# : 2005005(pdf)
Last modified : May 25, 2006
 
2Q/2006 Inspection Findings - Callaway                                                                                                      Page 1 of 6 Callaway 2Q/2006 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures Resulted in a Main Steam Line Water Hammer A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after a water hammer transient occurred because plant operators failed to follow a procedure. On May 31, 2006, a main steam line water hammer occurred after plant operators failed to properly align the main steam drains prior to initializing a reactor coolant system heat up. Plant operators had failed to return the drain valves to service following main turbine repairs. This issue was entered into the corrective action program as Callaway Action Request 200604255.
This finding is greater than minor because this finding is associated with the initiating events cornerstone configuration control attribute for equipment lineup in that it challenged one main steam line and the associated components upstream of the main steam isolation valves. The inspectors used the at-power significance determination process because plant operators had secured the residual heat removal pump at the time of the event. This finding is of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to the likelihood of a reactor trip or the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding.
This finding had a crosscutting aspect in the area of human performance because plant operators failed to follow established procedures.
Inspection Report# : 2006003(pdf)
Mitigating Systems Significance:        Jun 23, 2006 Identified By: Self-Revealing Item Type: FIN Finding An Inadequate Switchyard Restoration Procedure Resulted in a Partial Loss of Off-Site Power A self-revealing finding was identified after an inadequate switchyard maintenance procedure resulted in the loss of power to a safety-related bus.
On June 6, 2006, off-site power was lost to a plant safety-related bus when electricians restored the breaker failure relay for a main switchyard breaker. The emergency diesel generator automatically started and restored power to the bus. The inspectors identified AmerenUE did not use applicable operational experience prior to conducting the work. NRC Information Notice 1991-81, Switchyard Problems that Contribute to Loss of Offsite Power, and an AmerenUE operational experience, Lessons Learned Switchyard Activity Checklist, addressed similar conditions. This issue was entered into the corrective action program as Callaway Action Request 200604492.
This finding is greater than minor because the availability and reliability of a safety-related 4 kV bus was challenged. This finding was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the objective to ensure availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined this finding to be of very low safety significance because the condition was not a design or qualification deficiency per Part 9900, Technical Guidance, Operability Determination Process, did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of human performance because personnel did not have adequate procedures and work instructions for switchyard work.
Inspection Report# : 2006003(pdf)
Significance:        Apr 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Recognize and Correct Inadequate Emergency Procedures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to take adequate corrective action to prevent recurrence of a significant condition adverse to quality. Specifically, AmerenUE failed to correct the Emergency Operating Procedure deficiencies associated with Final Safety Analysis Report requirements following an April 15, 1998 notification of the same deficiencies at another standardized nuclear unit power plant system plant. At that time AmerenUE did not identify and correct similar deficiencies involving the component cooling water system support function for residual heat removal heat exchangers. The Emergency Operating Procedure deficiencies were discovered by plant personnel on March 27, 2006, during a simulator exercise involving the transition to the emergency core cooling system recirculation phase. Problem identification and resolution crosscutting aspects were identified for the failure to adequately identify and correct Emergency Operating Procedures deficiencies to ensure operation within the design basis.
 
2Q/2006 Inspection Findings - Callaway                                                                                                      Page 2 of 6 This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability. The failure to provide for component cooling water system flow through the residual heat removal heat exchangers for initial containment recirculation could result in a loss of the component cooling water system and thus become a much more significant safety concern. AmerenUEs evaluation of the condition was considered for the time allowable to establish component cooling water flow before a loss of the component cooling water system would occur.
AmerenUE provided an evaluation that demonstrated a loss of component cooling water would not occur based on the timing of operator actions.
Because the timing did affect the probabilistic risk assessment for human reliability, a Phase 3 risk assessment was performed by an NRC senior reactor analyst. The analyst determined that the finding was of very low safety significance, Green. AmerenUE entered this issue into their corrective action program as Callaway Action Request 200602565.
Inspection Report# : 2006011(pdf)
Significance:        Apr 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Result in Possible CCW Runout Conditions The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for AmerenUEs failure to implement appropriate corrective actions for maintaining component cooling water flow consistent with design basis requirements. On April 11 and 12, 2006, AmerenUE placed the Train A component cooling water system in a configuration which could result in component cooling water pump runout in the event of a loss-of-coolant accident coincident with a loss of offsite power. Crosscutting aspects associated with problem identification and resolution were identified for the failure to implement appropriate corrective actions to ensure the component cooling water system remained operable for other design basis events.
This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability in that a loss of one train of the component cooling water system could cause other mitigating equipment (i.e., pumps and heat exchangers) to fail and thus become a much more significant safety concern. Using the NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Screening Worksheet, the finding was determined to be of very low safety significance because it did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time. AmerenUE entered this issue into its corrective action program as Callaway Action Request 200602995.
Inspection Report# : 2006011(pdf)
Significance:        Mar 24, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of Degraded Plant Equipment The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," after the licensee failed to promptly identify, evaluate, and correct a degraded control building air conditioning unit compressor. The compressor developed a hole in one of the cylinder head discharge reed valves. The hole allowed the bypass of hot discharge gases and rendered the compressor incapable of completing the safety function for the specified mission time. The hole was caused by cyclic fatigue stress. This issue was entered into the corrective action program as Callaway Action Request 200601177. This finding is associated with the crosscutting area of problem identification and resolution because the issue involved the failure of operations personnel to adequately evaluate degraded plant equipment.
This finding is greater than minor because, if left uncorrected, the degradation would have worsened and become a more significant safety concern.
This finding was a qualification deficiency that resulted in loss of operability per "Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment." However, the finding is of very low safety significance because it did not represent a loss of system safety function, did not represent an actual loss of safety function for a single train for greater than the 30-day Technical Specification allowed outage time, did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk-significant per 10 CFR 50.65, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Minimum gap size exceeded for containment recirculation sump.
The inspectors identified a noncited violation of 10 CFR Part 50, Criterion X, after plant quality control personnel performed an inadequate inspection of an emergency core cooling system containment recirculation sump. The inspection failed to identify a 11/2-inch hole which provided a path for foreign material into the containment sump which could affect the recirculation mode of emergency core cooling system operation.
AmerenUE completed a detailed inspection of the sump on April 27, 2004 in response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors," but failed to identify the 11/2 -inch hole. This issue was entered into the corrective action program as Callaway Action Request 200509189.
This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure availability and reliability of the containment recirculation sump emergency core cooling system containment safety function. This finding is of very low safety significance because the condition was a qualification deficiency confirmed not to result in loss of function per Part 9900, Technical Assessment; "Operability Determination Process for Operability and Functional Assessment." The
 
2Q/2006 Inspection Findings - Callaway                                                                                                      Page 3 of 6 cause of this finding, poor attention to detail by personnel, is related to the crosscutting element of human performance.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately implement continuous compensatory fire watches.
The inspectors identified a noncited violation of Technical Specification 5.4.1, "Procedures," associated with seven examples of inadequately performed continuous fire watches. In September 2005, AmerenUE provided verbal guidance to fire watch personnel that continuous watches may be met by a 15 minute roving fire patrol. The roving patrol did not ensure adequate compensatory action for fire areas with degraded detection or suppression capability. As a result, fire watch personnel were not available to promptly detect, report, and extinguish a fire while still in the incipient stage. AmerenUE did not evaluate this change to ensure no adverse affect on the ability to achieve and maintain safe shutdown in the event a fire was created. The condition was entered into the corrective action program as Callaway Action Request 200510325. The cause of this finding is related to the crosscutting element of human performance because the resources needed to support the task, including complete and accurate procedures and supervision, were less than adequate.
This finding is greater than minor because inadequate fire watches are associated with the reactor safety mitigating systems cornerstone attribute to provide protection against external factors and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding is of very low safety significance because the condition had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for outages or degradations.
A low degradation rating was assigned to this finding as the provision affected by this finding is expected to display nearly the same level of effectiveness and reliability.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Conduct Simulator Testing in Accordance with ANSI/ANS 3.5-1998 The inspectors determined that the failure to adhere to ANSI/ANS 3.5-1998, as endorsed by Regulatory Guide 1.149 "Nuclear Power Plant Simulation Facilities for Use in Operator Training and License Examinations,"Revision 3, October 2001, as committed to in the Callaway Plant Simulation certification dated March 13, 2000, was a finding. Specifically, the simulator performance testing did not meet the standards specified in ANSI/ANS 3.5-1998, in that: (1) all required parameters during the simulator test were not recorded; and (2) simulator to baseline data comparisons were unavailable.
The failure to evaluate and document simulator performance testing is more than minor because it affected the Operator Requalification attribute of the Mitigating Systems and Initiating Event cornerstone of reactor safety and is inconsistent with the requirements of 10 CFR 55.46 in that simulator fidelity issues may not be identified, which have the potential of causing negative training. The finding was considered to be of very low safety significance because the discrepancies have not yet impacted operator actions in the plant, such that, safety-related equipment was made inoperable or that operators failed to properly respond to plant transients.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Use of a Non-Qualified Calculation in a Safety Related Modification The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," associated with an inadequate engineering procedure used for the verification of design calculations. The inadequate procedure resulted in a non-qualified, non-safety-related engineering calculation being used to demonstrate that the safety-related containment recirculation sump valves were capable of performing the safety function described in the design bases. The performance deficiency associated with this finding involved the failure of engineering personnel to only use qualified calculations for safety-related applications. The cause of this finding is related to the crosscutting element of human performance because insufficient resources were provided to ensure complete and accurate procedures to support task performance. This finding was entered into the Corrective Action Program as Callaway Action Request 200509849.
This finding is greater than minor because if left uncorrected, this finding would become a more significant safety concern. This finding is determined to have very low safety significance because this issue involves a design deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessment."
Inspection Report# : 2005005(pdf)
Significance:        Sep 23, 2005
 
2Q/2006 Inspection Findings - Callaway                                                                                                        Page 4 of 6 Identified By: Self-Revealing Item Type: NCV NonCited Violation Degraded auxiliary feedwater pump due to the failure to follow procedure.
A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after AmerenUE failed to properly align the turbine driven auxiliary feedwater pump mechanical overspeed trip mechanism after surveillance testing. The trip mechanism was misaligned from August 1 - 18, 2005. The misaligned trip mechanism increased the probability the turbine would trip if the pump would have been required to respond to an event. This issue was entered into the corrective action program as Callaway Action Request 200505801. This finding, which involved the failure of an operator to follow procedure, was associated with the crosscutting area of human performance.
This finding is greater than minor because the degraded trip mechanism affected the reactor mitigating systems cornerstone and the equipment performance attribute to ensure availability of systems that respond to prevent core damage. This finding is only of very low safety significance because the condition was not a design or qualification deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result in an actual loss of safety function of a system; did not increase the likelihood of a fire; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2005004(pdf)
Barrier Integrity Significance:        Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less Than Adequate Evaluation of Containment Heat Exchanger Postmodification Tests Results and self Assessment Recommendations The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, after containment heat exchanger postmodification tests, conducted in Refuel Outages 11 (May 2001) and 12 (November 2002), failed to demonstrate that the system would perform satisfactorily in service. The inspectors identified that postmodification tests did not meet acceptance criteria, testing was not performed under appropriate conditions, test methods did not meet industry standards, and tests did not establish complete acceptance criteria. This issue was entered into the corrective action program as Callaway Action Requests 200509450, 200600012, and 200605143.
This finding is greater than minor because it affects the barrier integrity cornerstone and if left uncorrected, this finding could become a more significant safety concern for maintaining functionality of the containment. The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, guidance because this finding involved an actual reduction in defense-in-depth for the atmospheric pressure control of containment. The inspectors determined that this finding was Type B because the integrity of containment was affected without increasing the likelihood of core damage. The finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms, but not large early release frequency.
Inspection Report# : 2006003(pdf)
Significance:        Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less Than Adequate Evaluation of Containment Heat Exchanger Performance Monitoring Requirements The inspectors identified a noncited violation of Technical Specification 3.6.6, Containment Spray and Cooling Systems, after AmerenUE failed to perform Surveillance Requirement 3.6.6.7 to verify minimum cooling water was provided to each containment cooling train between October 23, 2002, and June 26, 2006. Technical Specification Bases, Figure 3.6.6.7-1, Containment Cooler Heat Removal Minimum Cooling Flow Rates, provided an acceptable region for reduced service water flow as a function of the available fraction of rated heat exchanger heat removal capacity.
The acceptable region ensured sufficient duty to remove the required containment heat loads during accident conditions. AmerenUE had not performed adequate testing to determine the containment heat exchanger available percent of rated capacity. This issue was entered into the corrective action program as Callaway Action Request 200605143.
This finding is greater than minor because if left uncorrected, this finding could become a more significant safety concern. This finding affected the barrier integrity cornerstone for the heat removal capability of the containment cooling system. The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, because this finding involved an actual reduction in defense in depth for the atmospheric pressure control of the containment. The inspectors determined that this finding was Type B because the integrity of the containment was affected without increasing the likelihood of core damage. The inspectors concluded this finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms but not large early release frequency.
This finding had a crosscutting aspect in the area of problem identification and resolution because AmerenUE did not adequately evaluate containment heat exchanger problems such that the causes and extent of condition were properly classified, prioritized, and evaluated for operability and reportability.
Inspection Report# : 2006003(pdf)
Significance:        Jun 23, 2006 Identified By: NRC
 
2Q/2006 Inspection Findings - Callaway                                                                                                    Page 5 of 6 Item Type: NCV NonCited Violation Less than adequate Operability Determination of a Degraded Containment Heat Exchanger The inspectors identified a noncited violation of Technical Specification 3.6.6, Containment Spray and Cooling Systems, after AmerenUE failed to perform Surveillance Requirement 3.6.6.7 to verify minimum cooling water was provided to each containment cooling train between October 23, 2002, and June 26, 2006. Technical Specification Bases, Figure 3.6.6.7-1, Containment Cooler Heat Removal Minimum Cooling Flow Rates, provided an acceptable region for reduced service water flow as a function of the available fraction of rated heat exchanger heat removal capacity.
The acceptable region ensured sufficient duty to remove the required containment heat loads during accident conditions. AmerenUE had not performed adequate testing to determine the containment heat exchanger available percent of rated capacity. This issue was entered into the corrective action program as Callaway Action Request 200605143.
This finding is greater than minor because if left uncorrected, this finding could become a more significant safety concern. This finding affected the barrier integrity cornerstone for the heat removal capability of the containment cooling system. The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, because this finding involved an actual reduction in defense in depth for the atmospheric pressure control of the containment. The inspectors determined that this finding was Type B because the integrity of the containment was affected without increasing the likelihood of core damage. The inspectors concluded this finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms but not large early release frequency.
This finding had a crosscutting aspect in the area of problem identification and resolution because AmerenUE did not adequately evaluate containment heat exchanger problems such that the causes and extent of condition were properly classified, prioritized, and evaluated for operability and reportability.
Inspection Report# : 2006003(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulted in Violation of RCS Cooldown and Heatup Rate Limits.
The inspectors identified a noncited violation of Technical Specification 5.4.1.a, "Procedures," after AmerenUE Operations personnel failed to maintain the reactor coolant system temperature limits on two occasions. On November 7, 2005, plant operators decreased the reactor coolant system pressurizer surge line temperature 260 degrees Fahrenheit in a one-hour period. The operators conducted the rapid cooldown after several containment lead shield blanket polyvinylchloride covers left in containment melted. On November 8, 2005, plant operators increased the surge line temperature about 175 degrees Fahrenheit in a one-hour period. Plant Technical Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits,"
and Plant procedures required reactor coolant system component temperature changes (except the pressurizer) be limited to 100 degrees in one hour.
The cause of this finding is related to the crosscutting element of human performance because of personnel failure to follow procedures.
This finding was greater than minor because it is associated with the reactor safety barrier integrity cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure reasonable assurance that the reactor coolant system piping barrier will protect the public from radionuclide releases caused by accidents or events. This finding is determined to have very low safety significance because an engineering evaluation concluded that the temperature transient did not significantly increase the likelihood of a loss of reactor coolant system inventory or degrade the ability to terminate a leak path. This finding was placed in the Corrective Action Program as Callaway Action Requests 200509487 and 200509143.
Inspection Report# : 2005005(pdf)
Significance:        Sep 23, 2005 Identified By: NRC Item Type: NCV NonCited Violation Ineffective corrective actions resulted in degraded control building habitability boundary.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, "Corrective Action," after ineffective corrective actions resulted in a repeat degradation of a control building emergency ventilation habitability boundary door. AmerenUE's work control organization twice authorized work on the essential switchgear room to emergency diesel generator room door without approval of the shift operations department. As a result, shift operations did not understand that the habitability boundary had been compromised by the maintenance. This finding, which involved ineffective corrective actions to prevent the repeat degradation of the ventilation system habitability boundary door, was associated with the crosscutting area of problem identification and resolution.
This finding was greater than minor because it was associated with the integrity of the control building pressure envelope in that the degraded door would not meet its habitability function. The finding was only of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room.
Inspection Report# : 2005004(pdf)
Emergency Preparedness
 
2Q/2006 Inspection Findings - Callaway                                                                                                    Page 6 of 6 Occupational Radiation Safety Significance:        Oct 21, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to control a high radiation area with dose rates greater than 1.0 rem per hour.
The inspector reviewed a self-revealing non-cited violation of Technical Specification 5.7.2 because the licensee failed to control a high radiation area with dose rates greater than 1.0 rem per hour. Specifically, on September 26, 2005, the reactor vessel head was moved from the head stand and placed back on the reactor vessel without the proper radiological controls in place for a high radiation area with dose rates as high as 6.0 rem per hour. A loud noise created by the falling of a locking device on the reactor head alerted radiation protection personnel that the head lift had begun prematurely. The licensee's immediate corrective actions were to ensure that individuals were not present in the high radiation area and to place the reactor head in a safe condition on the reactor vessel. The finding was entered into the licensee's corrective action program as Callaway Action Request 200507546.
The failure to control a high radiation area with dose rates greater than 1.0 rem per hour is a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involved the potential for a worker's unplanned or unintended dose resulting from actions contrary to technical specifications. When processed through the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because the finding did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. In addition, this finding has crosscutting aspects associated with human performance because poor coordination and communication between the head lift crew and radiation protection personnel directly contributed to the finding.
Inspection Report# : 2005009(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Less Than Adequate Spent Fuel Pool Water Inventory Risk Controls The inspectors identified a finding after AmerenUE implemented less than adequate risk management controls of the spent fuel pool water inventory. On September 29, 2005, the core had been off-loaded to the spent fuel pool and the transfer canal weir was removed. The spent fuel pool temperature was 99 degrees Fahrenheit with a 12.1 hour time-to-boil. Transfer tube Valve ECV-995 isolated the fuel transfer canal from the containment cavity. In this configuration, the tube valve could provide a drain path reducing water level from 25 feet to less than 2 feet above the spent fuel. Valve ECV-995 was closed but was not identified in the shutdown risk management system and did not have administrative controls to protect against misalignment. NRC Information Notice 2005-16, "Outage Planning and Scheduling - Impacts on Risk," emphasized that most spent fuel pool events had a common thread of human error and involved equipment misalignment. This finding was entered into the Corrective Action Program as Callaway Action Requests 200507593 and 200507693.
This finding is greater than minor because if left uncorrected, it would have become a more significant safety concern. Because Manual Chapter 0609, "Significance Determination Process," does not specifically address findings related to the spent fuel pool inventory, this finding is determined to have very low safety significance based on NRC management review with input from senior reactor analysts. No violation of regulatory requirements occurred.
Inspection Report# : 2005005(pdf)
Last modified : August 25, 2006
 
3Q/2006 Inspection Findings - Callaway                                                                                Page 1 of 9 Callaway 3Q/2006 Plant Inspection Findings Initiating Events Significance:      Sep 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Equipment Control Procedur Resulted in Loss of Volume Control Tank Inventory A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified following two unplanned 50 gallon per minute volume control tank loss of inventory events. Both events occurred due to an inadequate equipment control procedure. On July 17 and 18, 2006, planned maintenance on the boron thermal regeneration system inlet valve created a flow path from the reactor coolant system letdown line to the equipment drain system from a known leaking demineralizer drain valve. AmerenUE did not have an administrative procedure or other effective means to control letdown line configuration with the leaking demineralizer drain valve. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200605751.
This finding is greater than minor because this finding is associated with the reactor safety initiating events cornerstone attribute of procedure quality and affected the objective to limit the likelihood of events that upset plant stability. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that this finding is only of very low significance because the condition did not result in the reactor coolant system Technical Specification leakage limit being exceeded (this leakage is not considered reactor coolant system leakage), did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or flooding. This finding has a crosscutting aspect in the area of human performance associated with resources because AmerenUE did not ensure a complete and accurate equipment control procedure was available to plant operators.
Inspection Report# : 2006004(pdf)
Significance:      Sep 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Review Adequacy of Procedure and Operator Response to a Turbine Trip A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after an inadequate turbine trip procedure resulted in an unplanned manual reactor trip. On May 12, 2006, the inadequate procedure lead to a steam generator level transient after plant operators failed to stabilize reactor power following a turbine trip. Operators manually tripped the reactor following a high steam generator level feedwater isolation. AmerenUE placed this issue in the corrective action program as Callaway Action Requests 200603734 and 200603736.
This finding is greater than minor because this finding is associated with the reactor safety initiating events cornerstone attributes of procedure quality and affects the objective to limit the likelihood of events that upset plant stability. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this finding to be of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding has a crosscutting aspect in the area of human performance associated with resources because AmerenUE did not ensure complete, accurate, up-to-date design documentation and procedures were available to plant operators.
Inspection Report# : 2006004(pdf)
Significance:      Sep 23, 2006 Identified By: NRC
 
3Q/2006 Inspection Findings - Callaway                                                                                Page 2 of 9 Item Type: FIN Finding Review of Less Than Adequate Post Reactor Trip Evaluation An NRC identified finding was identified after AmerenUE restarted the reactor on May 12, 2006, without completing an adequate reactor posttrip evaluation. The licensee did not adequately address discrepancies between expected and actual plant response during the transient leading to the reactor trip. The licensee did not identify the cause of the trip or implement immediate corrective actions prior to restart as required by plant administrative procedures. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200605766.
This finding is greater than minor because it could become a more significant event if left uncorrected. This finding is associated with the initiating events cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this finding is of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because AmerenUE did not thoroughly evaluate the cause of the reactor trip or implement timely corrective actions prior to the Emergency Duty Officer authorizing reactor restart.
Inspection Report# : 2006004(pdf)
Significance:        Jun 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures Resulted in a Main Steam Line Water Hammer A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after a water hammer transient occurred because plant operators failed to follow a procedure. On May 31, 2006, a main steam line water hammer occurred after plant operators failed to properly align the main steam drains prior to initializing a reactor coolant system heat up. Plant operators had failed to return the drain valves to service following main turbine repairs. This issue was entered into the corrective action program as Callaway Action Request 200604255.
This finding is greater than minor because this finding is associated with the initiating events cornerstone configuration control attribute for equipment lineup in that it challenged one main steam line and the associated components upstream of the main steam isolation valves. The inspectors used the at-power significance determination process because plant operators had secured the residual heat removal pump at the time of the event. This finding is of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to the likelihood of a reactor trip or the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding had a crosscutting aspect in the area of human performance because plant operators failed to follow established procedures.
Inspection Report# : 2006003(pdf)
Mitigating Systems Significance:        Jun 23, 2006 Identified By: Self-Revealing Item Type: FIN Finding An Inadequate Switchyard Restoration Procedure Resulted in a Partial Loss of Off-Site Power A self-revealing finding was identified after an inadequate switchyard maintenance procedure resulted in the loss of power to a safety-related bus. On June 6, 2006, off-site power was lost to a plant safety-related bus when electricians restored the breaker failure relay for a main switchyard breaker. The emergency diesel generator automatically started and restored power to the bus. The inspectors identified AmerenUE did not use applicable operational experience prior to conducting the work. NRC Information Notice 1991-81, Switchyard Problems that Contribute to Loss of Offsite Power, and an AmerenUE operational experience, Lessons Learned Switchyard Activity Checklist, addressed similar conditions. This issue was entered into the corrective action program as Callaway Action Request 200604492.
 
3Q/2006 Inspection Findings - Callaway                                                                                  Page 3 of 9 This finding is greater than minor because the availability and reliability of a safety-related 4 kV bus was challenged. This finding was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the objective to ensure availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined this finding to be of very low safety significance because the condition was not a design or qualification deficiency per Part 9900, Technical Guidance, Operability Determination Process, did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of human performance because personnel did not have adequate procedures and work instructions for switchyard work.
Inspection Report# : 2006003(pdf)
Significance:        Apr 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Recognize and Correct Inadequate Emergency Procedures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to take adequate corrective action to prevent recurrence of a significant condition adverse to quality. Specifically, AmerenUE failed to correct the Emergency Operating Procedure deficiencies associated with Final Safety Analysis Report requirements following an April 15, 1998 notification of the same deficiencies at another standardized nuclear unit power plant system plant. At that time AmerenUE did not identify and correct similar deficiencies involving the component cooling water system support function for residual heat removal heat exchangers. The Emergency Operating Procedure deficiencies were discovered by plant personnel on March 27, 2006, during a simulator exercise involving the transition to the emergency core cooling system recirculation phase. Problem identification and resolution crosscutting aspects were identified for the failure to adequately identify and correct Emergency Operating Procedures deficiencies to ensure operation within the design basis.
This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability.
The failure to provide for component cooling water system flow through the residual heat removal heat exchangers for initial containment recirculation could result in a loss of the component cooling water system and thus become a much more significant safety concern. AmerenUEs evaluation of the condition was considered for the time allowable to establish component cooling water flow before a loss of the component cooling water system would occur. AmerenUE provided an evaluation that demonstrated a loss of component cooling water would not occur based on the timing of operator actions.
Because the timing did affect the probabilistic risk assessment for human reliability, a Phase 3 risk assessment was performed by an NRC senior reactor analyst. The analyst determined that the finding was of very low safety significance, Green. AmerenUE entered this issue into their corrective action program as Callaway Action Request 200602565.
Inspection Report# : 2006011(pdf)
Significance:        Apr 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Result in Possible CCW Runout Conditions The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for AmerenUEs failure to implement appropriate corrective actions for maintaining component cooling water flow consistent with design basis requirements. On April 11 and 12, 2006, AmerenUE placed the Train A component cooling water system in a configuration which could result in component cooling water pump runout in the event of a loss-of-coolant accident coincident with a loss of offsite power. Crosscutting aspects associated with problem identification and resolution were identified for the failure to implement appropriate corrective actions to ensure the component cooling water system remained operable for other design basis events.
This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability in that a loss of one train of the component cooling water system could cause other mitigating equipment (i.e., pumps and heat exchangers) to fail and thus become a much more significant safety concern. Using the NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Screening Worksheet, the finding was determined to be of very low safety significance because it did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time. AmerenUE entered this issue into its corrective action program as Callaway Action
 
3Q/2006 Inspection Findings - Callaway                                                                              Page 4 of 9 Request 200602995.
Inspection Report# : 2006011(pdf)
Significance:        Mar 24, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of Degraded Plant Equipment The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," after the licensee failed to promptly identify, evaluate, and correct a degraded control building air conditioning unit compressor. The compressor developed a hole in one of the cylinder head discharge reed valves. The hole allowed the bypass of hot discharge gases and rendered the compressor incapable of completing the safety function for the specified mission time.
The hole was caused by cyclic fatigue stress. This issue was entered into the corrective action program as Callaway Action Request 200601177. This finding is associated with the crosscutting area of problem identification and resolution because the issue involved the failure of operations personnel to adequately evaluate degraded plant equipment.
This finding is greater than minor because, if left uncorrected, the degradation would have worsened and become a more significant safety concern. This finding was a qualification deficiency that resulted in loss of operability per "Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment." However, the finding is of very low safety significance because it did not represent a loss of system safety function, did not represent an actual loss of safety function for a single train for greater than the 30-day Technical Specification allowed outage time, did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk-significant per 10 CFR 50.65, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Minimum gap size exceeded for containment recirculation sump.
The inspectors identified a noncited violation of 10 CFR Part 50, Criterion X, after plant quality control personnel performed an inadequate inspection of an emergency core cooling system containment recirculation sump. The inspection failed to identify a 11/2-inch hole which provided a path for foreign material into the containment sump which could affect the recirculation mode of emergency core cooling system operation. AmerenUE completed a detailed inspection of the sump on April 27, 2004 in response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors," but failed to identify the 11/2 -inch hole. This issue was entered into the corrective action program as Callaway Action Request 200509189.
This finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure availability and reliability of the containment recirculation sump emergency core cooling system containment safety function. This finding is of very low safety significance because the condition was a qualification deficiency confirmed not to result in loss of function per Part 9900, Technical Assessment; "Operability Determination Process for Operability and Functional Assessment." The cause of this finding, poor attention to detail by personnel, is related to the crosscutting element of human performance.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately implement continuous compensatory fire watches.
The inspectors identified a noncited violation of Technical Specification 5.4.1, "Procedures," associated with seven examples of inadequately performed continuous fire watches. In September 2005, AmerenUE provided verbal guidance to
 
3Q/2006 Inspection Findings - Callaway                                                                                Page 5 of 9 fire watch personnel that continuous watches may be met by a 15 minute roving fire patrol. The roving patrol did not ensure adequate compensatory action for fire areas with degraded detection or suppression capability. As a result, fire watch personnel were not available to promptly detect, report, and extinguish a fire while still in the incipient stage.
AmerenUE did not evaluate this change to ensure no adverse affect on the ability to achieve and maintain safe shutdown in the event a fire was created. The condition was entered into the corrective action program as Callaway Action Request 200510325. The cause of this finding is related to the crosscutting element of human performance because the resources needed to support the task, including complete and accurate procedures and supervision, were less than adequate.
This finding is greater than minor because inadequate fire watches are associated with the reactor safety mitigating systems cornerstone attribute to provide protection against external factors and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
This finding is of very low safety significance because the condition had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for outages or degradations. A low degradation rating was assigned to this finding as the provision affected by this finding is expected to display nearly the same level of effectiveness and reliability.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Failure to Conduct Simulator Testing in Accordance with ANSI/ANS 3.5-1998 The inspectors determined that the failure to adhere to ANSI/ANS 3.5-1998, as endorsed by Regulatory Guide 1.149 "Nuclear Power Plant Simulation Facilities for Use in Operator Training and License Examinations,"Revision 3, October 2001, as committed to in the Callaway Plant Simulation certification dated March 13, 2000, was a finding. Specifically, the simulator performance testing did not meet the standards specified in ANSI/ANS 3.5-1998, in that: (1) all required parameters during the simulator test were not recorded; and (2) simulator to baseline data comparisons were unavailable.
The failure to evaluate and document simulator performance testing is more than minor because it affected the Operator Requalification attribute of the Mitigating Systems and Initiating Event cornerstone of reactor safety and is inconsistent with the requirements of 10 CFR 55.46 in that simulator fidelity issues may not be identified, which have the potential of causing negative training. The finding was considered to be of very low safety significance because the discrepancies have not yet impacted operator actions in the plant, such that, safety-related equipment was made inoperable or that operators failed to properly respond to plant transients.
Inspection Report# : 2005005(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Use of a Non-Qualified Calculation in a Safety Related Modification The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," associated with an inadequate engineering procedure used for the verification of design calculations. The inadequate procedure resulted in a non-qualified, non-safety-related engineering calculation being used to demonstrate that the safety-related containment recirculation sump valves were capable of performing the safety function described in the design bases. The performance deficiency associated with this finding involved the failure of engineering personnel to only use qualified calculations for safety-related applications. The cause of this finding is related to the crosscutting element of human performance because insufficient resources were provided to ensure complete and accurate procedures to support task performance. This finding was entered into the Corrective Action Program as Callaway Action Request 200509849.
This finding is greater than minor because if left uncorrected, this finding would become a more significant safety concern.
This finding is determined to have very low safety significance because this issue involves a design deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, "Operability Determination Process for Operability and Functional Assessment."
Inspection Report# : 2005005(pdf)
 
3Q/2006 Inspection Findings - Callaway                                                                                Page 6 of 9 Barrier Integrity Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less Than Adequate Evaluation of Containment Heat Exchanger Postmodification Tests Results and self Assessment Recommendations The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, after containment heat exchanger postmodification tests, conducted in Refuel Outages 11 (May 2001) and 12 (November 2002), failed to demonstrate that the system would perform satisfactorily in service. The inspectors identified that postmodification tests did not meet acceptance criteria, testing was not performed under appropriate conditions, test methods did not meet industry standards, and tests did not establish complete acceptance criteria. This issue was entered into the corrective action program as Callaway Action Requests 200509450, 200600012, and 200605143.
This finding is greater than minor because it affects the barrier integrity cornerstone and if left uncorrected, this finding could become a more significant safety concern for maintaining functionality of the containment. The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, guidance because this finding involved an actual reduction in defense-in-depth for the atmospheric pressure control of containment. The inspectors determined that this finding was Type B because the integrity of containment was affected without increasing the likelihood of core damage. The finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms, but not large early release frequency.
Inspection Report# : 2006003(pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less Than Adequate Evaluation of Containment Heat Exchanger Performance Monitoring Requirements The inspectors identified a noncited violation of Technical Specification 3.6.6, Containment Spray and Cooling Systems, after AmerenUE failed to perform Surveillance Requirement 3.6.6.7 to verify minimum cooling water was provided to each containment cooling train between October 23, 2002, and June 26, 2006. Technical Specification Bases, Figure 3.6.6.7-1, Containment Cooler Heat Removal Minimum Cooling Flow Rates, provided an acceptable region for reduced service water flow as a function of the available fraction of rated heat exchanger heat removal capacity. The acceptable region ensured sufficient duty to remove the required containment heat loads during accident conditions. AmerenUE had not performed adequate testing to determine the containment heat exchanger available percent of rated capacity. This issue was entered into the corrective action program as Callaway Action Request 200605143.
This finding is greater than minor because if left uncorrected, this finding could become a more significant safety concern.
This finding affected the barrier integrity cornerstone for the heat removal capability of the containment cooling system.
The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, because this finding involved an actual reduction in defense in depth for the atmospheric pressure control of the containment. The inspectors determined that this finding was Type B because the integrity of the containment was affected without increasing the likelihood of core damage. The inspectors concluded this finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms but not large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution because AmerenUE did not adequately evaluate containment heat exchanger problems such that the causes and extent of condition were properly classified, prioritized, and evaluated for operability and reportability.
Inspection Report# : 2006003(pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less than adequate Operability Determination of a Degraded Containment Heat Exchanger The inspectors identified a noncited violation of Technical Specification 3.6.6, Containment Spray and Cooling Systems,
 
3Q/2006 Inspection Findings - Callaway                                                                              Page 7 of 9 after AmerenUE failed to perform Surveillance Requirement 3.6.6.7 to verify minimum cooling water was provided to each containment cooling train between October 23, 2002, and June 26, 2006. Technical Specification Bases, Figure 3.6.6.7-1, Containment Cooler Heat Removal Minimum Cooling Flow Rates, provided an acceptable region for reduced service water flow as a function of the available fraction of rated heat exchanger heat removal capacity. The acceptable region ensured sufficient duty to remove the required containment heat loads during accident conditions. AmerenUE had not performed adequate testing to determine the containment heat exchanger available percent of rated capacity. This issue was entered into the corrective action program as Callaway Action Request 200605143.
This finding is greater than minor because if left uncorrected, this finding could become a more significant safety concern.
This finding affected the barrier integrity cornerstone for the heat removal capability of the containment cooling system.
The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, because this finding involved an actual reduction in defense in depth for the atmospheric pressure control of the containment. The inspectors determined that this finding was Type B because the integrity of the containment was affected without increasing the likelihood of core damage. The inspectors concluded this finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms but not large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution because AmerenUE did not adequately evaluate containment heat exchanger problems such that the causes and extent of condition were properly classified, prioritized, and evaluated for operability and reportability.
Inspection Report# : 2006003(pdf)
Significance:      Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulted in Violation of RCS Cooldown and Heatup Rate Limits.
The inspectors identified a noncited violation of Technical Specification 5.4.1.a, "Procedures," after AmerenUE Operations personnel failed to maintain the reactor coolant system temperature limits on two occasions. On November 7, 2005, plant operators decreased the reactor coolant system pressurizer surge line temperature 260 degrees Fahrenheit in a one-hour period. The operators conducted the rapid cooldown after several containment lead shield blanket polyvinylchloride covers left in containment melted. On November 8, 2005, plant operators increased the surge line temperature about 175 degrees Fahrenheit in a one-hour period. Plant Technical Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and Plant procedures required reactor coolant system component temperature changes (except the pressurizer) be limited to 100 degrees in one hour. The cause of this finding is related to the crosscutting element of human performance because of personnel failure to follow procedures.
This finding was greater than minor because it is associated with the reactor safety barrier integrity cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure reasonable assurance that the reactor coolant system piping barrier will protect the public from radionuclide releases caused by accidents or events. This finding is determined to have very low safety significance because an engineering evaluation concluded that the temperature transient did not significantly increase the likelihood of a loss of reactor coolant system inventory or degrade the ability to terminate a leak path. This finding was placed in the Corrective Action Program as Callaway Action Requests 200509487 and 200509143.
Inspection Report# : 2005005(pdf)
Emergency Preparedness Significance:      Sep 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Program Failure to Ensure Emergency Action Level Entered when Meeting the Defined Limit for Hazardous Atmosphere The inspectors identified a Green noncited violation of 10 CFR 50.54(q) for a failure to adequately implement the emergency plan. The licensee failed to declare an ALERT when conditions existed that met Emergency Action Level 3J,
 
3Q/2006 Inspection Findings - Callaway                                                                              Page 8 of 9 Hazards Affecting Plant Safety. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200607835.
This finding is greater than minor because this finding is associated with the reactor safety emergency preparedness cornerstone attribute of emergency response organization performance and affects the cornerstone objective of the licensee protecting public health and safety during a radiological emergency. The inspectors used Manual Chapter 0609, Significance Determination Process, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, because the licensee misunderstood the emergency action level, but otherwise adequately implemented the emergency plan. The inspectors concluded this finding is of very low safety significance because the performance deficiency is related to the inability to implement one emergency action level at the ALERT level, which is a risk significant planning standard problem but not a risk significant planning standard function failure or a risk significant planning standard degraded function. This finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not provide training to the emergency response organization that clearly communicated the basis for decisions associated with the language changes made to Emergency Action Level 3J.
Inspection Report# : 2006004(pdf)
Occupational Radiation Safety Significance:      Oct 21, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to control a high radiation area with dose rates greater than 1.0 rem per hour.
The inspector reviewed a self-revealing non-cited violation of Technical Specification 5.7.2 because the licensee failed to control a high radiation area with dose rates greater than 1.0 rem per hour. Specifically, on September 26, 2005, the reactor vessel head was moved from the head stand and placed back on the reactor vessel without the proper radiological controls in place for a high radiation area with dose rates as high as 6.0 rem per hour. A loud noise created by the falling of a locking device on the reactor head alerted radiation protection personnel that the head lift had begun prematurely. The licensee's immediate corrective actions were to ensure that individuals were not present in the high radiation area and to place the reactor head in a safe condition on the reactor vessel. The finding was entered into the licensee's corrective action program as Callaway Action Request 200507546.
The failure to control a high radiation area with dose rates greater than 1.0 rem per hour is a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involved the potential for a worker's unplanned or unintended dose resulting from actions contrary to technical specifications. When processed through the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because the finding did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. In addition, this finding has crosscutting aspects associated with human performance because poor coordination and communication between the head lift crew and radiation protection personnel directly contributed to the finding.
Inspection Report# : 2005009(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
 
3Q/2006 Inspection Findings - Callaway                                                                          Page 9 of 9 Miscellaneous Significance:      Dec 31, 2005 Identified By: NRC Item Type: FIN Finding Less Than Adequate Spent Fuel Pool Water Inventory Risk Controls The inspectors identified a finding after AmerenUE implemented less than adequate risk management controls of the spent fuel pool water inventory. On September 29, 2005, the core had been off-loaded to the spent fuel pool and the transfer canal weir was removed. The spent fuel pool temperature was 99 degrees Fahrenheit with a 12.1 hour time-to-boil. Transfer tube Valve ECV-995 isolated the fuel transfer canal from the containment cavity. In this configuration, the tube valve could provide a drain path reducing water level from 25 feet to less than 2 feet above the spent fuel. Valve ECV-995 was closed but was not identified in the shutdown risk management system and did not have administrative controls to protect against misalignment. NRC Information Notice 2005-16, "Outage Planning and Scheduling - Impacts on Risk," emphasized that most spent fuel pool events had a common thread of human error and involved equipment misalignment. This finding was entered into the Corrective Action Program as Callaway Action Requests 200507593 and 200507693.
This finding is greater than minor because if left uncorrected, it would have become a more significant safety concern.
Because Manual Chapter 0609, "Significance Determination Process," does not specifically address findings related to the spent fuel pool inventory, this finding is determined to have very low safety significance based on NRC management review with input from senior reactor analysts. No violation of regulatory requirements occurred.
Inspection Report# : 2005005(pdf)
Last modified : December 21, 2006
 
4Q/2006 Inspection Findings - Callaway                                                                                Page 1 of 9 Callaway 4Q/2006 Plant Inspection Findings Initiating Events Significance:      Sep 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Equipment Control Procedur Resulted in Loss of Volume Control Tank Inventory A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified following two unplanned 50 gallon per minute volume control tank loss of inventory events. Both events occurred due to an inadequate equipment control procedure. On July 17 and 18, 2006, planned maintenance on the boron thermal regeneration system inlet valve created a flow path from the reactor coolant system letdown line to the equipment drain system from a known leaking demineralizer drain valve. AmerenUE did not have an administrative procedure or other effective means to control letdown line configuration with the leaking demineralizer drain valve. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200605751.
This finding is greater than minor because this finding is associated with the reactor safety initiating events cornerstone attribute of procedure quality and affected the objective to limit the likelihood of events that upset plant stability. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that this finding is only of very low significance because the condition did not result in the reactor coolant system Technical Specification leakage limit being exceeded (this leakage is not considered reactor coolant system leakage), did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or flooding. This finding has a crosscutting aspect in the area of human performance associated with resources because AmerenUE did not ensure a complete and accurate equipment control procedure was available to plant operators.
Inspection Report# : 2006004 (pdf)
Significance:      Sep 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Review Adequacy of Procedure and Operator Response to a Turbine Trip A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after an inadequate turbine trip procedure resulted in an unplanned manual reactor trip. On May 12, 2006, the inadequate procedure lead to a steam generator level transient after plant operators failed to stabilize reactor power following a turbine trip. Operators manually tripped the reactor following a high steam generator level feedwater isolation. AmerenUE placed this issue in the corrective action program as Callaway Action Requests 200603734 and 200603736.
This finding is greater than minor because this finding is associated with the reactor safety initiating events cornerstone attributes of procedure quality and affects the objective to limit the likelihood of events that upset plant stability. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this finding to be of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding has a crosscutting aspect in the area of human performance associated with resources because AmerenUE did not ensure complete, accurate, up-to-date design documentation and procedures were available to plant operators.
Inspection Report# : 2006004 (pdf)
Significance:      Sep 23, 2006 Identified By: NRC
 
4Q/2006 Inspection Findings - Callaway                                                                                Page 2 of 9 Item Type: FIN Finding Review of Less Than Adequate Post Reactor Trip Evaluation An NRC identified finding was identified after AmerenUE restarted the reactor on May 12, 2006, without completing an adequate reactor posttrip evaluation. The licensee did not adequately address discrepancies between expected and actual plant response during the transient leading to the reactor trip. The licensee did not identify the cause of the trip or implement immediate corrective actions prior to restart as required by plant administrative procedures. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200605766.
This finding is greater than minor because it could become a more significant event if left uncorrected. This finding is associated with the initiating events cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this finding is of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because AmerenUE did not thoroughly evaluate the cause of the reactor trip or implement timely corrective actions prior to the Emergency Duty Officer authorizing reactor restart.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures Resulted in a Main Steam Line Water Hammer A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after a water hammer transient occurred because plant operators failed to follow a procedure. On May 31, 2006, a main steam line water hammer occurred after plant operators failed to properly align the main steam drains prior to initializing a reactor coolant system heat up. Plant operators had failed to return the drain valves to service following main turbine repairs. This issue was entered into the corrective action program as Callaway Action Request 200604255.
This finding is greater than minor because this finding is associated with the initiating events cornerstone configuration control attribute for equipment lineup in that it challenged one main steam line and the associated components upstream of the main steam isolation valves. The inspectors used the at-power significance determination process because plant operators had secured the residual heat removal pump at the time of the event. This finding is of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to the likelihood of a reactor trip or the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding had a crosscutting aspect in the area of human performance because plant operators failed to follow established procedures.
Inspection Report# : 2006003 (pdf)
Mitigating Systems Significance:        Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Callaway Action Request The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to initiate Callaway Action Requests for conditions adverse to quality that affected the reliability of mitigating systems. Specifically, on August 17, 2005, and on May 30, 2006, the licensee discovered a high point air trap in the Train A safety injection discharge piping and decreasing water level in Steam Generators A and D; however, the licensee failed to enter these conditions adverse to quality into their corrective action program. The water in the main steam line contributed to a water hammer and the void had the potential to impact operability of the safety injection system. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609812.
 
4Q/2006 Inspection Findings - Callaway                                                                                Page 3 of 9 The performance deficiency involved the failure to initiate corrective action documents for identified conditions adverse to quality, as required. This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the reliability and availability of systems that respond to initiating events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone, was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. The finding has cross-cutting aspects related to problem identification and resolution, in that, personnel did not identify issues at a low threshold and in a timely manner commensurate with their safety significance.
Inspection Report# : 2006012 (pdf)
Significance:      Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify conditions adverse to quality The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, and the corrective action program because licensee personnel failed to recognize and to identify two separate examples as conditions adverse to quality.
Specifically, on April 13, 2006, and on October 17, 2006, licensee personnel did not identify blocked containment cooler tubes and a dirty emergency diesel generator turbocharger air intake filter, respectively, as conditions adverse to quality.
Failure to recognize these conditions as degraded and identify them as conditions adverse to quality, delayed the immediate evaluation of operability and implementation of corrective actions. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609813.
The performance deficiency involved the failure to promptly identify and correct conditions adverse to quality. The inappropriate classification of Callaway Action Requests 200602989 and 200608806 as Action Notice Callaway Action Requests delayed and prevented actions required by the corrective action program. This finding is greater than minor because a later evaluation by the licensee determined that safety related equipment had been adversely affected. [This deficiency is similar to Manual Chapter 0612, Appendix E, Example 4.a.] Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. The finding has cross-cutting aspects related to problem identification and resolution, in that, personnel did not identify issues at a low threshold and in a timely manner commensurate with their safety significance.
Inspection Report# : 2006012 (pdf)
Significance:      Nov 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to effectively implement actions to prevent recurrence A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, resulted from the failure to correct, and preclude repetition of (evaluate extent of condition), a significant condition adverse to quality related to identification of high spots in horizontal safety injection system discharge piping. Specifically, the licensee failed to identify all high spots in the susceptible discharge piping in February 2005; consequently, a modification did not prevent recurrence of voids collecting in high spots. The licensee entered the deficiency into their corrective action program as Callaway Action Request 200608644.
The performance deficiency involved the failure to effectively evaluate all susceptible points in the Train A safety injection discharge piping. This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the availability of systems that respond to initiating events. The failure of the design change affected the reliability of the safety injection system. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a cross-cutting aspect related to problem identification and resolution, in that, the licensee did not thoroughly evaluate the voiding problems such that the resolutions addressed the extent of condition.
 
4Q/2006 Inspection Findings - Callaway                                                                              Page 4 of 9 Inspection Report# : 2006012 (pdf)
Significance:      Nov 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to promptly correct a condition adverse to quality.
A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI resulted after operations personnel failed to implement corrective actions. Specifically, the licensee failed to modify Procedure OSP-AL-V0003, Auxiliary Feedwater Pump Discharge Check Valve (ALV0054) Closure Test, to ensure that upstream piping would be vented prior to performing the test to prevent overpressurizing the turbine-driven auxiliary feedwater pump suction pipe. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200509277.
The performance deficiency involved the failure to change a procedure as recommended in a corrective action to prevent recurrence. This finding associated with failure to implement corrective action is greater than minor because, if left uncorrected, the finding would become a more significant safety concern. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with resources because the licensee did not ensure complete, accurate, up-to-date procedures were available to plant operators.
Inspection Report# : 2006012 (pdf)
Significance: N/A Oct 13, 2006 Identified By: NRC Item Type: FIN Finding Supplemental inspection following a white mitigating systems performance index heat removal system performance indicator.
The U.S. Nuclear Regulatory Commission performed this supplemental inspection to assess the licensees evaluation associated with a performance indicator (Mitigating Systems Performance Index Heat Removal System) that became White with the initial implementation of the Mitigating Systems Performance Index performance indicators during the second quarter of 2006. The primary reason for this performance indicator being characterized as White was system reliability for the auxiliary feedwater system. The licensee performed a comprehensive evaluation that identified three primary root causes for the degraded reliability of the auxiliary feedwater system: poor implementation of maintenance programs to improve quality; a lack of training for maintenance personnel; and poor coordination of personnel and resources. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee, in general, adequately determined the root and contributing causes of the White performance indicator and established appropriate corrective actions. In addition, the licensee conducted an extent of cause review, which included a performance assessment of the remaining mitigating systems.
Inspection Report# : 2006013 (pdf)
Significance:      Jun 23, 2006 Identified By: Self-Revealing Item Type: FIN Finding An Inadequate Switchyard Restoration Procedure Resulted in a Partial Loss of Off-Site Power A self-revealing finding was identified after an inadequate switchyard maintenance procedure resulted in the loss of power to a safety-related bus. On June 6, 2006, off-site power was lost to a plant safety-related bus when electricians restored the breaker failure relay for a main switchyard breaker. The emergency diesel generator automatically started and restored power to the bus. The inspectors identified AmerenUE did not use applicable operational experience prior to conducting the work. NRC Information Notice 1991-81, Switchyard Problems that Contribute to Loss of Offsite Power, and an AmerenUE operational experience, Lessons Learned Switchyard Activity Checklist, addressed similar conditions. This issue was entered into the corrective action program as Callaway Action Request 200604492.
This finding is greater than minor because the availability and reliability of a safety-related 4 kV bus was challenged. This finding was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the objective to ensure availability and reliability of systems that respond to initiating events to prevent undesirable
 
4Q/2006 Inspection Findings - Callaway                                                                                  Page 5 of 9 consequences. The inspectors determined this finding to be of very low safety significance because the condition was not a design or qualification deficiency per Part 9900, Technical Guidance, Operability Determination Process, did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of human performance because personnel did not have adequate procedures and work instructions for switchyard work.
Inspection Report# : 2006003 (pdf)
Significance:        Apr 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Recognize and Correct Inadequate Emergency Procedures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to take adequate corrective action to prevent recurrence of a significant condition adverse to quality. Specifically, AmerenUE failed to correct the Emergency Operating Procedure deficiencies associated with Final Safety Analysis Report requirements following an April 15, 1998 notification of the same deficiencies at another standardized nuclear unit power plant system plant. At that time AmerenUE did not identify and correct similar deficiencies involving the component cooling water system support function for residual heat removal heat exchangers. The Emergency Operating Procedure deficiencies were discovered by plant personnel on March 27, 2006, during a simulator exercise involving the transition to the emergency core cooling system recirculation phase. Problem identification and resolution crosscutting aspects were identified for the failure to adequately identify and correct Emergency Operating Procedures deficiencies to ensure operation within the design basis.
This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability.
The failure to provide for component cooling water system flow through the residual heat removal heat exchangers for initial containment recirculation could result in a loss of the component cooling water system and thus become a much more significant safety concern. AmerenUEs evaluation of the condition was considered for the time allowable to establish component cooling water flow before a loss of the component cooling water system would occur. AmerenUE provided an evaluation that demonstrated a loss of component cooling water would not occur based on the timing of operator actions.
Because the timing did affect the probabilistic risk assessment for human reliability, a Phase 3 risk assessment was performed by an NRC senior reactor analyst. The analyst determined that the finding was of very low safety significance, Green. AmerenUE entered this issue into their corrective action program as Callaway Action Request 200602565.
Inspection Report# : 2006011 (pdf)
Significance:        Apr 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Result in Possible CCW Runout Conditions The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for AmerenUEs failure to implement appropriate corrective actions for maintaining component cooling water flow consistent with design basis requirements. On April 11 and 12, 2006, AmerenUE placed the Train A component cooling water system in a configuration which could result in component cooling water pump runout in the event of a loss-of-coolant accident coincident with a loss of offsite power. Crosscutting aspects associated with problem identification and resolution were identified for the failure to implement appropriate corrective actions to ensure the component cooling water system remained operable for other design basis events.
This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability in that a loss of one train of the component cooling water system could cause other mitigating equipment (i.e., pumps and heat exchangers) to fail and thus become a much more significant safety concern. Using the NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Screening Worksheet, the finding was determined to be of very low safety significance because it did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time. AmerenUE entered this issue into its corrective action program as Callaway Action Request 200602995.
Inspection Report# : 2006011 (pdf)
 
4Q/2006 Inspection Findings - Callaway                                                                                  Page 6 of 9 Significance:      Mar 24, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of Degraded Plant Equipment The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," after the licensee failed to promptly identify, evaluate, and correct a degraded control building air conditioning unit compressor. The compressor developed a hole in one of the cylinder head discharge reed valves. The hole allowed the bypass of hot discharge gases and rendered the compressor incapable of completing the safety function for the specified mission time.
The hole was caused by cyclic fatigue stress. This issue was entered into the corrective action program as Callaway Action Request 200601177. This finding is associated with the crosscutting area of problem identification and resolution because the issue involved the failure of operations personnel to adequately evaluate degraded plant equipment.
This finding is greater than minor because, if left uncorrected, the degradation would have worsened and become a more significant safety concern. This finding was a qualification deficiency that resulted in loss of operability per "Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment." However, the finding is of very low safety significance because it did not represent a loss of system safety function, did not represent an actual loss of safety function for a single train for greater than the 30-day Technical Specification allowed outage time, did not represent an actual loss of safety function of one or more non-Technical Specification trains of equipment designated as risk-significant per 10 CFR 50.65, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2006002 (pdf)
Barrier Integrity Significance:      Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate operability determination of a degraded main steam isolation valve The team identified a noncited violation of Technical Specification 3.7.2, after operations personnel failed to enter and implement required Technical Specification 3.7.2 actions. Specifically, the licensee had performed an inadequate operability determination related to a degraded main steam isolation valve that resulted in exceeding the allowed Technical Specifications out-of-service time between December 29 and 31, 2004. On October 19, 2006, the NRC determined that the licensee should have declared the main steam isolation valve and its actuation channel inoperable after removing one of two hydraulic actuators from service. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609233.
The performance deficiency involved the failure to perform an adequate operability evaluation of degraded plant equipment. As a result, the licensee failed to comply with the Technical Specifications. This finding is greater than minor because the configuration control attribute of the barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events is affected. The team used the At Power Significance Determination Process, of Manual Chapter 0609. The team concluded that a Phase 2 analysis was required because this finding affects both the fuel and containment barriers.
The team performed a Phase 2 analysis using the Risk-Informed Inspection Notebook for Callaway Nuclear Generating Station Unit 1, Revision 2. The team assumed that (1) one of two actuator trains was unavailable on one main steam isolation valve for less than 3 days and (2) the degraded actuator did not reduce the remaining main steam isolation valve mitigation capability credit to less than full mitigation credit. Based on the results of the Phase 2 analysis, this finding is determined to have very low safety significance. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly and correctly evaluate the operability of the degraded main steam isolation valve.
Inspection Report# : 2006012 (pdf)
 
4Q/2006 Inspection Findings - Callaway                                                                                Page 7 of 9 Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less Than Adequate Evaluation of Containment Heat Exchanger Postmodification Tests Results and self Assessment Recommendations The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, after containment heat exchanger postmodification tests, conducted in Refuel Outages 11 (May 2001) and 12 (November 2002), failed to demonstrate that the system would perform satisfactorily in service. The inspectors identified that postmodification tests did not meet acceptance criteria, testing was not performed under appropriate conditions, test methods did not meet industry standards, and tests did not establish complete acceptance criteria. This issue was entered into the corrective action program as Callaway Action Requests 200509450, 200600012, and 200605143.
This finding is greater than minor because it affects the barrier integrity cornerstone and if left uncorrected, this finding could become a more significant safety concern for maintaining functionality of the containment. The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, guidance because this finding involved an actual reduction in defense-in-depth for the atmospheric pressure control of containment. The inspectors determined that this finding was Type B because the integrity of containment was affected without increasing the likelihood of core damage. The finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms, but not large early release frequency.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less Than Adequate Evaluation of Containment Heat Exchanger Performance Monitoring Requirements The inspectors identified a noncited violation of Technical Specification 3.6.6, Containment Spray and Cooling Systems, after AmerenUE failed to perform Surveillance Requirement 3.6.6.7 to verify minimum cooling water was provided to each containment cooling train between October 23, 2002, and June 26, 2006. Technical Specification Bases, Figure 3.6.6.7-1, Containment Cooler Heat Removal Minimum Cooling Flow Rates, provided an acceptable region for reduced service water flow as a function of the available fraction of rated heat exchanger heat removal capacity. The acceptable region ensured sufficient duty to remove the required containment heat loads during accident conditions. AmerenUE had not performed adequate testing to determine the containment heat exchanger available percent of rated capacity. This issue was entered into the corrective action program as Callaway Action Request 200605143.
This finding is greater than minor because if left uncorrected, this finding could become a more significant safety concern.
This finding affected the barrier integrity cornerstone for the heat removal capability of the containment cooling system.
The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, because this finding involved an actual reduction in defense in depth for the atmospheric pressure control of the containment. The inspectors determined that this finding was Type B because the integrity of the containment was affected without increasing the likelihood of core damage. The inspectors concluded this finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms but not large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution because AmerenUE did not adequately evaluate containment heat exchanger problems such that the causes and extent of condition were properly classified, prioritized, and evaluated for operability and reportability.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less than adequate Operability Determination of a Degraded Containment Heat Exchanger The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after AmerenUE failed to properly evaluate a degraded containment cooling train. The inspectors identified that between August 16 and September 17, 2005, the performance data for Containment Cooler Train A did not demonstrate that the cooler was capable of performing the required design bases function because of fouling. AmerenUE performed an inadequate
 
4Q/2006 Inspection Findings - Callaway                                                                              Page 8 of 9 evaluation before placing the degraded heat exchanger in service for an 18-month fuel cycle beginning June 12, 2004. This issue was entered into the corrective action program as Callaway Action Request 200600012.
This finding is greater than minor because it affected the barrier integrity cornerstone for the heat removal capability of the containment cooling system and if left uncorrected, this finding could become a more significant safety concern because significant degradation of the containment cooler was not predicted or detected prior to the end of the operating cycle. The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, because this finding involved an actual reduction in defense in depth for the atmospheric pressure control of the containment. The inspectors determined that this finding was Type B because the integrity of the containment was affected without increasing the likelihood of core damage. The inspectors concluded this finding was of very low safety significance because the containment cooler heat exchanger only impacted late containment failure and source terms but not large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution because AmerenUE did not adequately evaluate operability of a degraded containment heat exchanger such that the resolutions addressed causes and extent of condition, as necessary.
Inspection Report# : 2006003 (pdf)
Emergency Preparedness Significance:      Sep 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Program Failure to Ensure Emergency Action Level Entered when Meeting the Defined Limit for Hazardous Atmosphere The inspectors identified a Green noncited violation of 10 CFR 50.54(q) for a failure to adequately implement the emergency plan. The licensee failed to declare an ALERT when conditions existed that met Emergency Action Level 3J, Hazards Affecting Plant Safety. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200607835.
This finding is greater than minor because this finding is associated with the reactor safety emergency preparedness cornerstone attribute of emergency response organization performance and affects the cornerstone objective of the licensee protecting public health and safety during a radiological emergency. The inspectors used Manual Chapter 0609, Significance Determination Process, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, because the licensee misunderstood the emergency action level, but otherwise adequately implemented the emergency plan. The inspectors concluded this finding is of very low safety significance because the performance deficiency is related to the inability to implement one emergency action level at the ALERT level, which is a risk significant planning standard problem but not a risk significant planning standard function failure or a risk significant planning standard degraded function. This finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not provide training to the emergency response organization that clearly communicated the basis for decisions associated with the language changes made to Emergency Action Level 3J.
Inspection Report# : 2006004 (pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection
 
4Q/2006 Inspection Findings - Callaway                                                                              Page 9 of 9 Physical Protection information not publicly available.
Miscellaneous Significance: N/A Nov 03, 2006 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed 230 Callaway Action Requests, several job orders, engineering evaluations, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities.
The team concluded that, generally, the licensee effectively identified, evaluated and prioritized, and implemented effective corrective actions for conditions adverse to quality. However, the team identified that additional effort is needed in all three areas. The team identified some instances of failure to initiate corrective action documents and numerous examples of failure to appropriately classify deficiencies as conditions adverse to quality. The team determined that quality and documentation for operability assessments has not improved significantly over the course of the evaluation period. Further, on occasion personnel were not self-critical as reflected by poor operational decision making. Two examples of findings reflect the condition of the corrective action problem evaluation activities in the mid portion of the assessment period. The team remained concerned that a lack of understanding of the detailed design and licensing basis continued to be evident in problem resolution. The team concluded that the licensee, generally, implemented timely, effective corrective actions, although some examples indicate continuing weakness in this area.
The team determined that the licensee had increased efforts to evaluate existing industry operating experience for relevance to the facility, and had entered identified items in the corrective action program; however, the team identified some examples that contributed to plant events.
The extensive performance improvement plan developed to address the substantive cross-cutting issue in human performance has addressed daily worker practice issues very well, although recent events occurred that indicate challenges remain. The increased management involvement in the corrective action program and in daily activities assisted in the improved performance. The team determined that licensee audits and assessments became more detailed, probing and self-critical with better assessments at the end of the assessment period. The licensee used benchmarking of industry best practices and third party evaluations that improved the corrective action program during this assessment period. While some of the changes were too recent to evaluate, the team concluded that improvements in the significant root cause process, Corrective Action Review Board graded approach, and scope and timing of corrective actions had improved.
On the basis of formal and informal interviews conducted during this inspection, the team determined that employees will raise issues to their supervision, use the corrective action program, and if necessary, bring concerns to the employee concerns program. The team concluded that the licensee established an acceptable and improving safety-conscious work environment. However, some indication exists that additional effort is needed to encourage the free flow of information to ensure safety issues are resolved promptly.
Inspection Report# : 2006012 (pdf)
Last modified : March 01, 2007
 
Callaway 1Q/2007 Plant Inspection Findings Initiating Events Significance:      Mar 24, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Management of an Operator Workaround Resulted in Unplanned Loss of Voume Control Tank Inventory The inspectors identified a finding after volume control tank inventory was inadvertently diverted from the reactor coolant system due to inadequate management of an operator workaround. On January 19 and March 22, 2007, operators had isolated the volume control tank from the demineralizer during resin transfer operations. However, volume control tank inventory was lost due to leakage past closed demineralizer isolation valves. Degraded Grinnell diaphragm valves have been a longstanding Callaway Plant material condition problem. Plant operations did not track nor effectively work around the degraded demineralizer valves.
This finding is greater than minor because the failure to adequately manage operator workarounds could reasonably be viewed as a precursor to a significant event. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because the condition did not result in the reactor coolant system technical specification leakage limit being exceeded, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or internal/external flood. This finding has a crosscutting aspect in the area of human performance associated with the work control component because AmerenUE did not plan work activities to support long-term equipment reliability by limiting operator workarounds. The licensee entered this finding into their corrective action program as Callaway Action Request 200700517.
Inspection Report# : 2007002 (pdf)
Significance:      Sep 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Equipment Control Procedur Resulted in Loss of Volume Control Tank Inventory A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified following two unplanned 50 gallon per minute volume control tank loss of inventory events. Both events occurred due to an inadequate equipment control procedure. On July 17 and 18, 2006, planned maintenance on the boron thermal regeneration system inlet valve created a flow path from the reactor coolant system letdown line to the equipment drain system from a known leaking demineralizer drain valve. AmerenUE did not have an administrative procedure or other effective means to control letdown line configuration with the leaking demineralizer drain valve. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200605751.
This finding is greater than minor because this finding is associated with the reactor safety initiating events cornerstone attribute of procedure quality and affected the objective to limit the likelihood of events that upset plant stability. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that this finding is only of very low significance because the condition did not result in the reactor coolant system Technical Specification leakage limit being exceeded (this leakage is not considered reactor coolant system leakage), did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or flooding. This finding has a crosscutting aspect in the area of human performance associated with resources because AmerenUE did not ensure a complete and accurate equipment control procedure was available to plant operators.
Inspection Report# : 2006004 (pdf)
 
Significance:        Sep 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Review Adequacy of Procedure and Operator Response to a Turbine Trip A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after an inadequate turbine trip procedure resulted in an unplanned manual reactor trip. On May 12, 2006, the inadequate procedure lead to a steam generator level transient after plant operators failed to stabilize reactor power following a turbine trip. Operators manually tripped the reactor following a high steam generator level feedwater isolation. AmerenUE placed this issue in the corrective action program as Callaway Action Requests 200603734 and 200603736.
This finding is greater than minor because this finding is associated with the reactor safety initiating events cornerstone attributes of procedure quality and affects the objective to limit the likelihood of events that upset plant stability. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this finding to be of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding has a crosscutting aspect in the area of human performance associated with resources because AmerenUE did not ensure complete, accurate, up-to-date design documentation and procedures were available to plant operators.
Inspection Report# : 2006004 (pdf)
Significance:        Sep 23, 2006 Identified By: NRC Item Type: FIN Finding Review of Less Than Adequate Post Reactor Trip Evaluation An NRC identified finding was identified after AmerenUE restarted the reactor on May 12, 2006, without completing an adequate reactor posttrip evaluation. The licensee did not adequately address discrepancies between expected and actual plant response during the transient leading to the reactor trip. The licensee did not identify the cause of the trip or implement immediate corrective actions prior to restart as required by plant administrative procedures. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200605766.
This finding is greater than minor because it could become a more significant event if left uncorrected. This finding is associated with the initiating events cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this finding is of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because AmerenUE did not thoroughly evaluate the cause of the reactor trip or implement timely corrective actions prior to the Emergency Duty Officer authorizing reactor restart.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures Resulted in a Main Steam Line Water Hammer A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after a water hammer transient occurred because plant operators failed to follow a procedure. On May 31, 2006, a main steam line water hammer occurred after plant operators failed to properly align the main steam drains prior to initializing a reactor coolant system heat up. Plant operators had failed to return the drain valves to service following main turbine repairs. This issue was entered into the corrective action program as Callaway Action Request 200604255.
This finding is greater than minor because this finding is associated with the initiating events cornerstone configuration control attribute for equipment lineup in that it challenged one main steam line and the associated components upstream of the main steam isolation valves. The inspectors used the at-power significance determination process because plant operators had secured the residual heat removal pump at the time of the event. This finding is of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to the likelihood of a
 
reactor trip or the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding had a crosscutting aspect in the area of human performance because plant operators failed to follow established procedures.
Inspection Report# : 2006003 (pdf)
Mitigating Systems Significance:        Mar 24, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inoperable Auxiliary Feedwater Pump due to an Inadequae Sureveillance Procedure A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an inadequate surveillance procedure resulted in the inadvertent defeat of the Train B turbine-driven auxiliary feedwater pump automatic start feature and an unplanned actuation of a cross-train control room ventilation isolation. On February 12, 2007, plant instrumentation and control technicians were performing a control room ventilation response time test. The procedure required the operator to block a high radiation test signal. The operator was unable to locate the block switch. A control room supervisor authorized a change to the procedure, which resulted in an incorrect block switch being used. The control room supervisor failed to verify correct block switch identification prior to authorizing the surveillance procedure change.
This finding is greater than minor because the failure to use an adequate surveillance procedure is associated with the mitigating systems cornerstone attribute of procedure quality and affects the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because it was not a design or qualification deficiency, did not result in loss-of-safety function of a single train for greater than the technical specifications allowed outage time, and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the control room supervisor did not thoroughly evaluate the apparent procedure problem before approving the change. This issue was entered into the licensee's corrective action program as Callaway Action Request 200701336.
Inspection Report# : 2007002 (pdf)
Significance:        Mar 24, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions to Preserve Essential Service Water System Material Condition A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an inadequate surveillance procedure resulted in the inadvertent defeat of the Train B turbine-driven auxiliary feedwater pump automatic start feature and an unplanned actuation of a cross-train control room ventilation isolation. On February 12, 2007, plant instrumentation and control technicians were performing a control room ventilation response time test. The procedure required the operator to block a high radiation test signal. The operator was unable to locate the block switch. A control room supervisor authorized a change to the procedure, which resulted in an incorrect block switch being used. The control room supervisor failed to verify correct block switch identification prior to authorizing the surveillance procedure change.
This finding is greater than minor because the failure to use an adequate surveillance procedure is associated with the mitigating systems cornerstone attribute of procedure quality and affects the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because it was not a design or qualification deficiency, did not result in loss-of-safety function of a single train for greater than the technical specifications allowed outage time, and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the control room supervisor did not thoroughly evaluate the apparent procedure problem before approving the change. This issue was entered into the licensee's corrective action program as Callaway Action Request 200701336.
 
Inspection Report# : 2007002 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Refueling Water Storage Tank Vent Sizing Calculation The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for an inadequate refeuling water storage tank vent sizing calculation. The calculation assumed that only one low head safety injection pump would operate when it should have assumed that all six emergency core cooling and containment spray pumps would take suction from the tank. When corrected, the revised calculation resulted in reducing the allowable vent blockage area from approximately 68 percent to 30 percent. In response to the teams concerns, the licensee inspected the vent and found a small mesh screen on the vents exterior, which reduced the available design margin to approximately 5 percent.
Subsequently, the licensee performed a new finite element analysis to demonstrate that sufficient margin existed to account for screen blockage scenarios, such as freezing rain. The licensee has entered this finding into their corrective action program as Callaway Action Requests 200610359 and 200700115.
The failure to meet design control requirements associated with the refeuling water storage tank vent design was a performance deficiency. This finding is more than minor because it affected the mitigating system cornerstone objective (design control attribute) to ensure the reliability and capability of the equipment needed to mitigate initiating events. The finding also affected the barrier integrity cornerstone objective (design control attribute) of providing physical design barriers, such as containment, to protect the public from radio nuclide releases caused by accidents or events. The team used the Manual Chapter 0609, Significance Determination Process Phase 1 screening worksheet and determined that the finding required a Phase 2 significance determination because it impacted two different cornerstones (mitigating systems and barrier integrity). The team performed a Phase 2 significance determination and determined that the finding was of very low safety significance. Only the large break loss-of-coolant accident sequence was affected. In addition, the safety injection and containment spray systems remained available.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Diesel Generator Fuel Oil Verification The team identified a noncited violation of Technical Specifications Surveillance Requirement 3.8.3.3 for the failure to verify that fuel oil testing results were within the specified limits. Consequently, fuel oil that was transferred to the Train A storage tank in October 2005 was out of specification for cetane and no actions were taken to evaluate or otherwise address the concern until identified by the NRC. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700100.
The failure to follow plant technical specifications and properly verify that the cetane level of new fuel oil was within the limits of the Diesel Fuel Oil Testing Program was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (human performance attribute) of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This finding had a crosscutting aspect in the area of human performance (work practices attribute), in that the chemistry technician failed to use appropriate self-checking work practices when verifying the sample results.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Diesel Generator Heat Exchanger Tube Plugging Calculation The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly calculate the tube plugging limit for the emergency diesel generator intercooler, jacket water, and lube oil cooler
 
heat exchangers. The calculation determined that approximately 1/3 of the tubes could be plugged without challenging emergency diesel generator operability under worst case design basis conditions. When corrected, the revised calculation resulted in reducing the allowable number of plugged tubes by approximately 40 percent. The licensee has entered this finding into their corrective action program as Callaway Action Requests 200700063 and 200700096.
The failure to implement appropriate design controls for safety-related tube plugging calculations was a performance deficiency. This finding is more than minor because it affected the mitigating system cornerstone objective (Design Control) to ensure the reliability and capability of the equipment needed to mitigate initiating events. In addition, the finding was more that minor because, if left uncorrected, it could result in a more significant safety concern. Specifically, if the heat exchanger tubes were plugged to the limit the heat exchangers may be inoperable under certain design basis conditions (i.e., higher essential service water temperatures). Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Essential Service Water Cooling Tower Design Basis Information into Specifications and Procedures.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly translate design requirements into procedures and instructions. Specifically, the cooling tower sizing calculation specified that a flow rate of 15,000 gallons per minute was necessary to meet design basis accident needs but flow balance procedures only required a flow rate of 11,724 gallons per minute. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700218.
The team determined that the failure to properly translate design information (essential service water flow rate through the cooling tower) into specifications and procedures was a performance deficiency. This finding was more than minor because it affected the mitigating system cornerstone objective (Procedure Quality Attribute) to ensure the reliability and capability of the equipment needed to mitigate initiating events. Further, if left uncorrected, it could lead to a more significant issue.
Specifically, information from the calculation could be used in other design documents and operability determinations.
Over-predicting cooling tower capability could mask other operational issues. Using the Manual Chapter 0609, Phase 1 screening worksheet, the team determined that the finding had very low safety significance (Green) because the finding was a design deficiency confirmed not to result in loss of operability in accordance with Part 9900 Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Initiate an Operability Evaluation for Water Hammer Concerns.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Procedures, for the failure to follow Callaway Plant procedure requirements associated with operability determinations. Specifically, engineers had identified that a water hammer was causing two residual heat removal system relief valves to fail and that the water hammer would likely recur in certain situations. The engineers failed to take the procedurally required actions to initiate a formal operability determination to evaluate the potential impact to the residual heat removal system pressure boundary. The licensee has entered this finding into their corrective action program as Callaway Action Request 200609805.
The failure to follow a Callaway Plant procedure was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Equipment Performance Attribute) of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This finding had a crosscutting aspect in the area of problem identification and resolution (corrective action program
 
component), in that engineers failed to performed the necessary proceduralized corrective actions to ensure that operability was properly evaluated.
Inspection Report# : 2006009 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify a Degraded Fire Barrier On November 21, 2006, the inspectors identified a noncited violation of Technical Specification 5.4.1.d, Fire Protection Program, after AmerenUE failed to identify and correct a degraded auxiliary building fire door. The inspectors identified that the latching mechanism on Fire Door 15031 would not engage because the double door had not been pinned. Failure of the door to latch resulted in a reduction in fire confinement capability. The door was required to provide a 3-hour fire barrier. The licensee had several prior opportunities to identify the degraded fire door. Security and operations personnel passed through the door several times each shift. The inspectors previously identified that the latch on Fire Door 15031 was degraded. Following the previous finding, AmerenUE implemented actions to increase the sensitivity of plant personnel to degraded fire doors. These actions were not effective to ensure that licensee personnel would recognize and enter the degraded fire door into the Corrective Action Program.
This finding is greater than minor because the degraded fire barrier affected the mitigating systems cornerstone external factors attribute objective to prevent undesirable consequences due to fire. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined this finding is in the fire confinement category and that the barrier was moderately degraded because the door latch was not functional. This finding is of very low safety significance because the exposed fire area contained no potential damage targets that are unique from those in the exposing fire area. The inspectors concluded that this finding has a problem identification and resolution crosscutting aspect associated with the corrective action program component because the licensee did not implement the corrective action program with a low threshold to identify the degraded door. The licensee entered this issue into the Corrective Action Program as Callaway Action Request CAR 20060962.
Inspection Report# : 2006005 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Categorize a Maintenance Preventable Functional Failure The inspectors identified a noncited violation of 10 CFR 50.65(a)(2) after AmerenUE failed to categorize the failure of motor-operated valve auxiliary contacts as a maintenance preventable functional failure and to monitor the component as required by 10 CFR 50.65(a)(1). On May 22, 2006, safety injection system motor-operated Valve EMHV8814A failed to open during surveillance testing due to stuck auxiliary contacts. On June 29, 2006, the Callaway maintenance rule expert panel concluded the failure was not a maintenance preventable functional failure. The inspectors reviewed the maintenance history of station motor-operated valves and determined eighteen previous auxiliary contact failures had occurred since 2002. Also, AmerenUE had initiated a modification to compensate for motor-operated valve electrical cubicle obsolescence and corrective action to address auxiliary contact failures. The inspectors determined that the June 29, 2006, expert panel incorrectly concluded that the auxiliary contact failures were not maintenance preventable. AmerenUE failed to perform an evaluation as required by 10 CFR 50.65(a)(1). On November 16, 2006 the expert panel reevaluated the failure of Valve EMHV8814A and five other auxiliary contact failures and concluded the failures were maintenance preventable functional failures and placed the auxiliary contacts system in 10 CFR 50.65(a)(1).
This finding is greater than minor because the failure of the expert panel to perform adequate evaluations would become a more significant safety concern if left uncorrected. This issue is similar to Example 7.b provided in Manual Chapter 0612, Appendix E. The inspectors analyzed this finding using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet. The inspectors determined this finding is of very low safety significance because, this finding is not a design or qualification deficiency, did not result in loss of safety function of a single train for greater than the allowed Technical Specification outage time and was not related to a seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the expert panel did not thoroughly or adequately evaluate the failure of the valve to address the causes and extent of condition. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200609603.
 
Inspection Report# : 2006005 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage Increased Risk During a Maintenance Activity On September 26, 2006, the inspectors identified a noncited violation of 10 CFR 50.65(a)(4) after AmerenUE failed to adequately manage the risk associated with maintenance on the turbine-driven auxiliary feedwater pump. AmerenUE removed the turbine-driven auxiliary feedwater pump from service for planned maintenance. The licensee determined this activity increased plant risk into the next higher risk configuration (Yellow). Procedure APA-ZZ-00315, Configuration Risk Management Program, required AmerenUE to take actions to protect redundant/diverse safety systems and components. Procedure APA-ZZ-00315 also stated that, if work could result in a risk-significant configuration or loss of system functions, consider use of physical barriers, such as ropes and/or signs to protect redundant/diverse components.
AmerenUE did not take adequate protective actions or use physical barriers on the redundant Train B motor-driven auxiliary feedwater pump. Plant workers passing through the motor-driven auxiliary feedwater pump room inadvertently rendered the pump inoperable by disabling the room cooler. The licensee determined that disabling the room cooler increased plant risk into the next higher risk configuration (Orange).
This finding is greater than minor because the licensee failed to implement prescribed significant compensatory measures during planned maintenance activity. This finding is similar to Example 7.g. provided in Manual Chapter 0612, Appendix E, because the auxiliary feedwater system key safety function was degraded. The inspectors used Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 2, Assessment of Risk Management Actions, to analyze this finding. The inspectors calculated an incremental core damage probability of 6.8 x 10-8 for the event, based a one-hour risk exposure duration and an increase of core damage probability from 1.8 x 10-4 to 7.1 x 10-4 after the Train B motor-driven auxiliary feedwater pump inadvertently rendered the pump inoperable. The inspectors determined the finding is of very low safety significance because incremental core damage probability 6.8 x 10-8 was less than 1.0 x 10-6. This finding has a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to appropriately plan work activities by incorporating risk insights and compensatory actions. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 20070284.
Inspection Report# : 2006005 (pdf)
Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Prior NRC Approval Before Removing Technical Specifications Limiting Condition for Operations On October 6, 2006, the inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 after AmerenUE failed to obtain prior NRC approval before removing the steam generator blowdown valve Limiting Condition for Operations requirement from the facility Technical Specifications. Part 50.36 of Title 10 of the Code of Federal Regulations, Technical Specifications, required AmerenUE to establish a Limiting Condition for Operations for components that are required to mitigate a design basis accident. The Callaway Plant accident analysis required the steam generator blowdown valves close to mitigate the steam line break accident and to ensure the auxiliary feedwater system safety function. AmerenUE met this requirement by including the blowdown valves in Technical Specification 3.6.3, Containment Isolation Valves, as referenced in FSAR Table 16.6-1, Containment Isolation Valves. On May 10, 2006, AmerenUE implemented FSAR Change Notice 02-012 which removed the blowdown valves from Table 16.6-1. This change removed the blowdown valves from within the scope the Technical Specifications Limiting Condition for Operations. The 50.59 safety evaluation supporting Change Notice 02-012 failed to identify that removal of the blowdown valves involved a change to the plant Technical Specifications and required prior NRC approval.
This issue involved traditional enforcement because AmerenUE did not receive prior NRC approval before changing the facility Technical Specifications. The inspectors evaluated this issue using Manual Chapter 0612, Appendix B. This issue is more than minor because the mitigating systems cornerstone attribute of equipment performance, reliability, and capability is impacted based on removal of the blowdown valve out-of-service time limits from the Technical Specifications. The inspectors used Manual Chapter 0609, Significance Determination Process, Phase 1, to analyze the safety significance of the violation. The inspectors concluded that the violation is of very low safety significance because the issue was not a design or qualification deficiency confirmed to result in loss of operability, did not represent a loss of system safety
 
function or an actual loss of safety function of one or more non-Technical Specification risk-significant equipment trains, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The issue has a problem identification and resolution crosscutting aspect associated with the corrective action program because the licensee's safety evaluation did not thoroughly evaluate the change such that the resolutions address causes and extent of conditions, as necessary. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200608902.
Inspection Report# : 2006005 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of an Operator Workaround Resulted in an Inoperable Safety Injection Accumulator The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after the licensee failed to adequately evaluate and promptly correct an operator workaround that resulted in the loss of nitrogen pressure on a safety injection accumulator. On December 3, 2006, Accumulator D was rendered inoperable due to low pressure. The low pressure condition occurred as plant operators attempted to add nitrogen to the accumulator. Plant operator efforts to work around degraded containment isolation and pressure relief valves during the filling operation resulted in an inoperable accumulator. The accumulator pressure had dropped below the minimum allowed Technical Specification pressure of 602 psig.
This issue is greater than minor because this finding is associated with the reactor safety mitigating systems cornerstone attribute of equipment performance and affects the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the inspectors determined that this finding is of very low significance because, although the condition did involve the loss of operability, it did not result in a loss of system safety train or function, and did not involve a seismic, flooding or severe weather event. This finding, which involved an inadequate evaluation of an operator workaround, has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because AmerenUE did not thoroughly evaluate problems such that resolutions addressed the causes and extent of conditions, as necessary. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200700286.
Inspection Report# : 2006005 (pdf)
Significance:      Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Callaway Action Request The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to initiate Callaway Action Requests for conditions adverse to quality that affected the reliability of mitigating systems. Specifically, on August 17, 2005, and on May 30, 2006, the licensee discovered a high point air trap in the Train A safety injection discharge piping and decreasing water level in Steam Generators A and D; however, the licensee failed to enter these conditions adverse to quality into their corrective action program. The water in the main steam line contributed to a water hammer and the void had the potential to impact operability of the safety injection system. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609812.
The performance deficiency involved the failure to initiate corrective action documents for identified conditions adverse to quality, as required. This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the reliability and availability of systems that respond to initiating events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone, was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. The finding has cross-cutting aspects related to problem identification and resolution, in that, personnel did not identify issues at a low threshold and in a timely manner commensurate with their safety significance.
Inspection Report# : 2006012 (pdf)
 
Significance:      Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify conditions adverse to quality The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, and the corrective action program because licensee personnel failed to recognize and to identify two separate examples as conditions adverse to quality.
Specifically, on April 13, 2006, and on October 17, 2006, licensee personnel did not identify blocked containment cooler tubes and a dirty emergency diesel generator turbocharger air intake filter, respectively, as conditions adverse to quality.
Failure to recognize these conditions as degraded and identify them as conditions adverse to quality, delayed the immediate evaluation of operability and implementation of corrective actions. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609813.
The performance deficiency involved the failure to promptly identify and correct conditions adverse to quality. The inappropriate classification of Callaway Action Requests 200602989 and 200608806 as Action Notice Callaway Action Requests delayed and prevented actions required by the corrective action program. This finding is greater than minor because a later evaluation by the licensee determined that safety related equipment had been adversely affected. [This deficiency is similar to Manual Chapter 0612, Appendix E, Example 4.a.] Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. The finding has cross-cutting aspects related to problem identification and resolution, in that, personnel did not identify issues at a low threshold and in a timely manner commensurate with their safety significance.
Inspection Report# : 2006012 (pdf)
Significance:      Nov 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to effectively implement actions to prevent recurrence A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, resulted from the failure to correct, and preclude repetition of (evaluate extent of condition), a significant condition adverse to quality related to identification of high spots in horizontal safety injection system discharge piping. Specifically, the licensee failed to identify all high spots in the susceptible discharge piping in February 2005; consequently, a modification did not prevent recurrence of voids collecting in high spots. The licensee entered the deficiency into their corrective action program as Callaway Action Request 200608644.
The performance deficiency involved the failure to effectively evaluate all susceptible points in the Train A safety injection discharge piping. This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the availability of systems that respond to initiating events. The failure of the design change affected the reliability of the safety injection system. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a cross-cutting aspect related to problem identification and resolution, in that, the licensee did not thoroughly evaluate the voiding problems such that the resolutions addressed the extent of condition.
Inspection Report# : 2006012 (pdf)
Significance:      Nov 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to promptly correct a condition adverse to quality.
A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI resulted after operations personnel failed to implement corrective actions. Specifically, the licensee failed to modify Procedure OSP-AL-V0003, Auxiliary Feedwater Pump Discharge Check Valve (ALV0054) Closure Test, to ensure that upstream piping would be vented prior to performing the test to prevent overpressurizing the turbine-driven auxiliary feedwater pump suction pipe. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200509277.
 
The performance deficiency involved the failure to change a procedure as recommended in a corrective action to prevent recurrence. This finding associated with failure to implement corrective action is greater than minor because, if left uncorrected, the finding would become a more significant safety concern. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with resources because the licensee did not ensure complete, accurate, up-to-date procedures were available to plant operators.
Inspection Report# : 2006012 (pdf)
Significance: N/A Oct 13, 2006 Identified By: NRC Item Type: FIN Finding Supplemental inspection following a white mitigating systems performance index heat removal system performance indicator.
The U.S. Nuclear Regulatory Commission performed this supplemental inspection to assess the licensees evaluation associated with a performance indicator (Mitigating Systems Performance Index Heat Removal System) that became White with the initial implementation of the Mitigating Systems Performance Index performance indicators during the second quarter of 2006. The primary reason for this performance indicator being characterized as White was system reliability for the auxiliary feedwater system. The licensee performed a comprehensive evaluation that identified three primary root causes for the degraded reliability of the auxiliary feedwater system: poor implementation of maintenance programs to improve quality; a lack of training for maintenance personnel; and poor coordination of personnel and resources. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee, in general, adequately determined the root and contributing causes of the White performance indicator and established appropriate corrective actions. In addition, the licensee conducted an extent of cause review, which included a performance assessment of the remaining mitigating systems.
Inspection Report# : 2006013 (pdf)
Significance:        Jun 23, 2006 Identified By: Self-Revealing Item Type: FIN Finding An Inadequate Switchyard Restoration Procedure Resulted in a Partial Loss of Off-Site Power A self-revealing finding was identified after an inadequate switchyard maintenance procedure resulted in the loss of power to a safety-related bus. On June 6, 2006, off-site power was lost to a plant safety-related bus when electricians restored the breaker failure relay for a main switchyard breaker. The emergency diesel generator automatically started and restored power to the bus. The inspectors identified AmerenUE did not use applicable operational experience prior to conducting the work. NRC Information Notice 1991-81, Switchyard Problems that Contribute to Loss of Offsite Power, and an AmerenUE operational experience, Lessons Learned Switchyard Activity Checklist, addressed similar conditions. This issue was entered into the corrective action program as Callaway Action Request 200604492.
This finding is greater than minor because the availability and reliability of a safety-related 4 kV bus was challenged. This finding was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the objective to ensure availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined this finding to be of very low safety significance because the condition was not a design or qualification deficiency per Part 9900, Technical Guidance, Operability Determination Process, did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of human performance because personnel did not have adequate procedures and work instructions for switchyard work.
Inspection Report# : 2006003 (pdf)
Significance:        Apr 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Recognize and Correct Inadequate Emergency Procedures
 
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to take adequate corrective action to prevent recurrence of a significant condition adverse to quality. Specifically, AmerenUE failed to correct the Emergency Operating Procedure deficiencies associated with Final Safety Analysis Report requirements following an April 15, 1998 notification of the same deficiencies at another standardized nuclear unit power plant system plant. At that time AmerenUE did not identify and correct similar deficiencies involving the component cooling water system support function for residual heat removal heat exchangers. The Emergency Operating Procedure deficiencies were discovered by plant personnel on March 27, 2006, during a simulator exercise involving the transition to the emergency core cooling system recirculation phase. Problem identification and resolution crosscutting aspects were identified for the failure to adequately identify and correct Emergency Operating Procedures deficiencies to ensure operation within the design basis.
This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability.
The failure to provide for component cooling water system flow through the residual heat removal heat exchangers for initial containment recirculation could result in a loss of the component cooling water system and thus become a much more significant safety concern. AmerenUEs evaluation of the condition was considered for the time allowable to establish component cooling water flow before a loss of the component cooling water system would occur. AmerenUE provided an evaluation that demonstrated a loss of component cooling water would not occur based on the timing of operator actions. Because the timing did affect the probabilistic risk assessment for human reliability, a Phase 3 risk assessment was performed by an NRC senior reactor analyst. The analyst determined that the finding was of very low safety significance, Green. AmerenUE entered this issue into their corrective action program as Callaway Action Request 200602565.
Inspection Report# : 2006011 (pdf)
Significance:        Apr 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions Result in Possible CCW Runout Conditions The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for AmerenUEs failure to implement appropriate corrective actions for maintaining component cooling water flow consistent with design basis requirements. On April 11 and 12, 2006, AmerenUE placed the Train A component cooling water system in a configuration which could result in component cooling water pump runout in the event of a loss-of-coolant accident coincident with a loss of offsite power. Crosscutting aspects associated with problem identification and resolution were identified for the failure to implement appropriate corrective actions to ensure the component cooling water system remained operable for other design basis events.
This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability in that a loss of one train of the component cooling water system could cause other mitigating equipment (i.e., pumps and heat exchangers) to fail and thus become a much more significant safety concern. Using the NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Screening Worksheet, the finding was determined to be of very low safety significance because it did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time. AmerenUE entered this issue into its corrective action program as Callaway Action Request 200602995.
Inspection Report# : 2006011 (pdf)
Barrier Integrity Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action for Refueling Water Storage Tank Vortexing Concerns The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action) for the failure to take adequate corrective actions following the identification of a condition adverse to quality. Specifically, the licensee had identified, in part, that a safety-related refueling water storage tank sizing calculation had failed to consider vortexing
 
at the tank suction inlet piping. This phenomena can cause air entrainment in pumps, which can lead to pump failure. The corrective measures were inadequate because engineers inappropriately used the margin associated with instrument uncertainty as if it were available design margin. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700224.
The team determined that the failure to take effective corrective measures to address a condition adverse to quality (failure to address vortexing in the refueling water storage tank sizing calculation) was a performance deficiency. The finding was more than minor because it affected the barrier integrity cornerstone objective (design control attribute) to provide reasonable assurance that physical design barriers (including the containment) protect the public from radio nuclide releases caused by accidents or events. The finding had crosscutting aspects in the area of problem identification and resolution (Operating Experience Attribute), in that the licensee had failed to adequately address the industry operating experience.
Inspection Report# : 2006009 (pdf)
Significance:        Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate operability determination of a degraded main steam isolation valve The team identified a noncited violation of Technical Specification 3.7.2, after operations personnel failed to enter and implement required Technical Specification 3.7.2 actions. Specifically, the licensee had performed an inadequate operability determination related to a degraded main steam isolation valve that resulted in exceeding the allowed Technical Specifications out-of-service time between December 29 and 31, 2004. On October 19, 2006, the NRC determined that the licensee should have declared the main steam isolation valve and its actuation channel inoperable after removing one of two hydraulic actuators from service. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609233.
The performance deficiency involved the failure to perform an adequate operability evaluation of degraded plant equipment. As a result, the licensee failed to comply with the Technical Specifications. This finding is greater than minor because the configuration control attribute of the barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events is affected. The team used the At Power Significance Determination Process, of Manual Chapter 0609. The team concluded that a Phase 2 analysis was required because this finding affects both the fuel and containment barriers.
The team performed a Phase 2 analysis using the Risk-Informed Inspection Notebook for Callaway Nuclear Generating Station Unit 1, Revision 2. The team assumed that (1) one of two actuator trains was unavailable on one main steam isolation valve for less than 3 days and (2) the degraded actuator did not reduce the remaining main steam isolation valve mitigation capability credit to less than full mitigation credit. Based on the results of the Phase 2 analysis, this finding is determined to have very low safety significance. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly and correctly evaluate the operability of the degraded main steam isolation valve.
Inspection Report# : 2006012 (pdf)
Significance:        Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less Than Adequate Evaluation of Containment Heat Exchanger Postmodification Tests Results and self Assessment Recommendations The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, after containment heat exchanger postmodification tests, conducted in Refuel Outages 11 (May 2001) and 12 (November 2002), failed to demonstrate that the system would perform satisfactorily in service. The inspectors identified that postmodification tests did not meet acceptance criteria, testing was not performed under appropriate conditions, test methods did not meet industry standards, and tests did not establish complete acceptance criteria. This issue was entered into the corrective action program as Callaway Action Requests 200509450, 200600012, and 200605143.
This finding is greater than minor because it affects the barrier integrity cornerstone and if left uncorrected, this finding could become a more significant safety concern for maintaining functionality of the containment. The inspectors used the
 
Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, guidance because this finding involved an actual reduction in defense-in-depth for the atmospheric pressure control of containment. The inspectors determined that this finding was Type B because the integrity of containment was affected without increasing the likelihood of core damage. The finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms, but not large early release frequency.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less Than Adequate Evaluation of Containment Heat Exchanger Performance Monitoring Requirements The inspectors identified a noncited violation of Technical Specification 3.6.6, Containment Spray and Cooling Systems, after AmerenUE failed to perform Surveillance Requirement 3.6.6.7 to verify minimum cooling water was provided to each containment cooling train between October 23, 2002, and June 26, 2006. Technical Specification Bases, Figure 3.6.6.7-1, Containment Cooler Heat Removal Minimum Cooling Flow Rates, provided an acceptable region for reduced service water flow as a function of the available fraction of rated heat exchanger heat removal capacity. The acceptable region ensured sufficient duty to remove the required containment heat loads during accident conditions.
AmerenUE had not performed adequate testing to determine the containment heat exchanger available percent of rated capacity. This issue was entered into the corrective action program as Callaway Action Request 200605143.
This finding is greater than minor because if left uncorrected, this finding could become a more significant safety concern.
This finding affected the barrier integrity cornerstone for the heat removal capability of the containment cooling system.
The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, because this finding involved an actual reduction in defense in depth for the atmospheric pressure control of the containment. The inspectors determined that this finding was Type B because the integrity of the containment was affected without increasing the likelihood of core damage. The inspectors concluded this finding was of very low safety significance because the containment heat exchanger only impacted late containment failure and source terms but not large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution because AmerenUE did not adequately evaluate containment heat exchanger problems such that the causes and extent of condition were properly classified, prioritized, and evaluated for operability and reportability.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Less than adequate Operability Determination of a Degraded Containment Heat Exchanger The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after AmerenUE failed to properly evaluate a degraded containment cooling train. The inspectors identified that between August 16 and September 17, 2005, the performance data for Containment Cooler Train A did not demonstrate that the cooler was capable of performing the required design bases function because of fouling. AmerenUE performed an inadequate evaluation before placing the degraded heat exchanger in service for an 18-month fuel cycle beginning June 12, 2004. This issue was entered into the corrective action program as Callaway Action Request 200600012.
This finding is greater than minor because it affected the barrier integrity cornerstone for the heat removal capability of the containment cooling system and if left uncorrected, this finding could become a more significant safety concern because significant degradation of the containment cooler was not predicted or detected prior to the end of the operating cycle. The inspectors used the Containment Integrity Significance Determination Process, Manual Chapter 0609, Appendix H, because this finding involved an actual reduction in defense in depth for the atmospheric pressure control of the containment. The inspectors determined that this finding was Type B because the integrity of the containment was affected without increasing the likelihood of core damage. The inspectors concluded this finding was of very low safety significance because the containment cooler heat exchanger only impacted late containment failure and source terms but not large early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution because AmerenUE did not adequately evaluate operability of a degraded containment heat exchanger such that the resolutions addressed causes and extent of condition, as necessary.
Inspection Report# : 2006003 (pdf)
 
Emergency Preparedness Significance:      Sep 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Program Failure to Ensure Emergency Action Level Entered when Meeting the Defined Limit for Hazardous Atmosphere The inspectors identified a Green noncited violation of 10 CFR 50.54(q) for a failure to adequately implement the emergency plan. The licensee failed to declare an ALERT when conditions existed that met Emergency Action Level 3J, Hazards Affecting Plant Safety. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200607835.
This finding is greater than minor because this finding is associated with the reactor safety emergency preparedness cornerstone attribute of emergency response organization performance and affects the cornerstone objective of the licensee protecting public health and safety during a radiological emergency. The inspectors used Manual Chapter 0609, Significance Determination Process, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, because the licensee misunderstood the emergency action level, but otherwise adequately implemented the emergency plan. The inspectors concluded this finding is of very low safety significance because the performance deficiency is related to the inability to implement one emergency action level at the ALERT level, which is a risk significant planning standard problem but not a risk significant planning standard function failure or a risk significant planning standard degraded function. This finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not provide training to the emergency response organization that clearly communicated the basis for decisions associated with the language changes made to Emergency Action Level 3J.
Inspection Report# : 2006004 (pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Nov 03, 2006 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed 230 Callaway Action Requests, several job orders, engineering evaluations, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities.
The team concluded that, generally, the licensee effectively identified, evaluated and prioritized, and implemented effective corrective actions for conditions adverse to quality. However, the team identified that additional effort is needed in all three areas. The team identified some instances of failure to initiate corrective action documents and numerous examples of
 
failure to appropriately classify deficiencies as conditions adverse to quality. The team determined that quality and documentation for operability assessments has not improved significantly over the course of the evaluation period. Further, on occasion personnel were not self-critical as reflected by poor operational decision making. Two examples of findings reflect the condition of the corrective action problem evaluation activities in the mid portion of the assessment period. The team remained concerned that a lack of understanding of the detailed design and licensing basis continued to be evident in problem resolution. The team concluded that the licensee, generally, implemented timely, effective corrective actions, although some examples indicate continuing weakness in this area.
The team determined that the licensee had increased efforts to evaluate existing industry operating experience for relevance to the facility, and had entered identified items in the corrective action program; however, the team identified some examples that contributed to plant events.
The extensive performance improvement plan developed to address the substantive cross-cutting issue in human performance has addressed daily worker practice issues very well, although recent events occurred that indicate challenges remain. The increased management involvement in the corrective action program and in daily activities assisted in the improved performance. The team determined that licensee audits and assessments became more detailed, probing and self-critical with better assessments at the end of the assessment period. The licensee used benchmarking of industry best practices and third party evaluations that improved the corrective action program during this assessment period. While some of the changes were too recent to evaluate, the team concluded that improvements in the significant root cause process, Corrective Action Review Board graded approach, and scope and timing of corrective actions had improved.
On the basis of formal and informal interviews conducted during this inspection, the team determined that employees will raise issues to their supervision, use the corrective action program, and if necessary, bring concerns to the employee concerns program. The team concluded that the licensee established an acceptable and improving safety-conscious work environment. However, some indication exists that additional effort is needed to encourage the free flow of information to ensure safety issues are resolved promptly.
Inspection Report# : 2006012 (pdf)
Last modified : June 01, 2007
 
Callaway 2Q/2007 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Management of an Operator Workaround Resulted in Unplanned Loss of Voume Control Tank Inventory The inspectors identified a finding after volume control tank inventory was inadvertently diverted from the reactor coolant system due to inadequate management of an operator workaround. On January 19 and March 22, 2007, operators had isolated the volume control tank from the demineralizer during resin transfer operations. However, volume control tank inventory was lost due to leakage past closed demineralizer isolation valves. Degraded Grinnell diaphragm valves have been a longstanding Callaway Plant material condition problem. Plant operations did not track nor effectively work around the degraded demineralizer valves.
This finding is greater than minor because the failure to adequately manage operator workarounds could reasonably be viewed as a precursor to a significant event. Using the Manual Chapter 0609, "Significance Determination Process,"
Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because the condition did not result in the reactor coolant system technical specification leakage limit being exceeded, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or internal/external flood. This finding has a crosscutting aspect in the area of human performance associated with the work control component because AmerenUE did not plan work activities to support long-term equipment reliability by limiting operator workarounds. The licensee entered this finding into their corrective action program as Callaway Action Request 200700517.
Inspection Report# : 2007002 (pdf)
Significance:        Sep 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Equipment Control Procedur Resulted in Loss of Volume Control Tank Inventory A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified following two unplanned 50 gallon per minute volume control tank loss of inventory events. Both events occurred due to an inadequate equipment control procedure. On July 17 and 18, 2006, planned maintenance on the boron thermal regeneration system inlet valve created a flow path from the reactor coolant system letdown line to the equipment drain system from a known leaking demineralizer drain valve. AmerenUE did not have an administrative procedure or other effective means to control letdown line configuration with the leaking demineralizer drain valve. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200605751.
This finding is greater than minor because this finding is associated with the reactor safety initiating events cornerstone attribute of procedure quality and affected the objective to limit the likelihood of events that upset plant stability. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that this finding is only of very low significance because the condition did not result in the reactor coolant system Technical Specification leakage limit being exceeded (this leakage is not considered reactor coolant system leakage), did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or flooding. This finding has a crosscutting aspect in the area of human performance associated with resources because AmerenUE did not ensure a complete and accurate equipment control procedure was available to plant operators.
Inspection Report# : 2006004 (pdf)
 
Significance:        Sep 23, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Review Adequacy of Procedure and Operator Response to a Turbine Trip A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after an inadequate turbine trip procedure resulted in an unplanned manual reactor trip. On May 12, 2006, the inadequate procedure lead to a steam generator level transient after plant operators failed to stabilize reactor power following a turbine trip. Operators manually tripped the reactor following a high steam generator level feedwater isolation.
AmerenUE placed this issue in the corrective action program as Callaway Action Requests 200603734 and 200603736.
This finding is greater than minor because this finding is associated with the reactor safety initiating events cornerstone attributes of procedure quality and affects the objective to limit the likelihood of events that upset plant stability. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this finding to be of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding has a crosscutting aspect in the area of human performance associated with resources because AmerenUE did not ensure complete, accurate, up-to-date design documentation and procedures were available to plant operators.
Inspection Report# : 2006004 (pdf)
Significance:        Sep 23, 2006 Identified By: NRC Item Type: FIN Finding Review of Less Than Adequate Post Reactor Trip Evaluation An NRC identified finding was identified after AmerenUE restarted the reactor on May 12, 2006, without completing an adequate reactor posttrip evaluation. The licensee did not adequately address discrepancies between expected and actual plant response during the transient leading to the reactor trip. The licensee did not identify the cause of the trip or implement immediate corrective actions prior to restart as required by plant administrative procedures. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200605766.
This finding is greater than minor because it could become a more significant event if left uncorrected. This finding is associated with the initiating events cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined this finding is of very low safety significance because the condition was not a loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would be unavailable, and did not increase the likelihood of fire or flooding. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because AmerenUE did not thoroughly evaluate the cause of the reactor trip or implement timely corrective actions prior to the Emergency Duty Officer authorizing reactor restart.
Inspection Report# : 2006004 (pdf)
Mitigating Systems Significance:        Mar 24, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inoperable Auxiliary Feedwater Pump due to an Inadequae Sureveillance Procedure A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an inadequate surveillance procedure resulted in the inadvertent defeat of the Train B turbine-driven auxiliary feedwater pump automatic start feature and an unplanned actuation of a cross-train control room ventilation isolation. On February 12, 2007, plant instrumentation and control technicians were performing a control room ventilation response time test.
The procedure required the operator to block a high radiation test signal. The operator was unable to locate the block
 
switch. A control room supervisor authorized a change to the procedure, which resulted in an incorrect block switch being used. The control room supervisor failed to verify correct block switch identification prior to authorizing the surveillance procedure change.
This finding is greater than minor because the failure to use an adequate surveillance procedure is associated with the mitigating systems cornerstone attribute of procedure quality and affects the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because it was not a design or qualification deficiency, did not result in loss-of-safety function of a single train for greater than the technical specifications allowed outage time, and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the control room supervisor did not thoroughly evaluate the apparent procedure problem before approving the change.
This issue was entered into the licensee's corrective action program as Callaway Action Request 200701336.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 24, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions to Preserve Essential Service Water System Material Condition The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria XVI, after past corrective actions were inadequate to preclude recurrence of essential service water piping degradation due to corrosion. On March 14 and 23, 2007, plant personnel identified through-wall leaks in the Train B large bore essential service water piping. Plant operators declared the essential service water train inoperable and implemented elevated plant risk and required implementation of risk management actions in both cases. Plant technicians performed non-destructive examinations on about 10 percent of the accessible large bore piping. Technicians identified 93 indications of less than minimum pipe wall thickness. The licensee concluded the pipe degradation resulted from microbiologically influenced corrosion. Poor material condition of the essential service water system has been a longstanding problem at the Callaway Plant. On March 23, 2005, plant personnel identified an essential service water through-wall leak in large bore piping, which required a technical specification required shutdown and on January 25, 2006, plant operators declared Train B of the essential service water system inoperable due to a through-wall pipe leak. These conditions were identified as significant conditions adverse to quality in the licensees corrective action program. The licensee's extent of condition review and corrective actions following the March 23, 2005, and January 25, 2006, occurrences were not adequate to prevent further examples of degraded essential service water piping from microbiologically influenced corrosion.
This finding is greater than minor because it is associated with the reactor safety mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, this finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because AmerenUE did not fully evaluate essential service water corrosion issues to ensure that the resolutions adequately addressed the causes and extent of condition needed to ensure nuclear safety. This issue was entered into the licensee's corrective action program as Callaway Action Request 200702724.
Inspection Report# : 2007002 (pdf)
Significance:      Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Refueling Water Storage Tank Vent Sizing Calculation The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for an inadequate refeuling water storage tank vent sizing calculation. The calculation assumed that only one low head safety injection pump would operate when it should have assumed that all six emergency core cooling and containment spray pumps would take suction from the tank. When corrected, the revised calculation resulted in reducing the
 
allowable vent blockage area from approximately 68 percent to 30 percent. In response to the teams concerns, the licensee inspected the vent and found a small mesh screen on the vents exterior, which reduced the available design margin to approximately 5 percent. Subsequently, the licensee performed a new finite element analysis to demonstrate that sufficient margin existed to account for screen blockage scenarios, such as freezing rain. The licensee has entered this finding into their corrective action program as Callaway Action Requests 200610359 and 200700115.
The failure to meet design control requirements associated with the refeuling water storage tank vent design was a performance deficiency. This finding is more than minor because it affected the mitigating system cornerstone objective (design control attribute) to ensure the reliability and capability of the equipment needed to mitigate initiating events. The finding also affected the barrier integrity cornerstone objective (design control attribute) of providing physical design barriers, such as containment, to protect the public from radio nuclide releases caused by accidents or events. The team used the Manual Chapter 0609, Significance Determination Process Phase 1 screening worksheet and determined that the finding required a Phase 2 significance determination because it impacted two different cornerstones (mitigating systems and barrier integrity). The team performed a Phase 2 significance determination and determined that the finding was of very low safety significance. Only the large break loss-of-coolant accident sequence was affected. In addition, the safety injection and containment spray systems remained available.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Diesel Generator Fuel Oil Verification The team identified a noncited violation of Technical Specifications Surveillance Requirement 3.8.3.3 for the failure to verify that fuel oil testing results were within the specified limits. Consequently, fuel oil that was transferred to the Train A storage tank in October 2005 was out of specification for cetane and no actions were taken to evaluate or otherwise address the concern until identified by the NRC. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700100.
The failure to follow plant technical specifications and properly verify that the cetane level of new fuel oil was within the limits of the Diesel Fuel Oil Testing Program was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (human performance attribute) of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This finding had a crosscutting aspect in the area of human performance (work practices attribute), in that the chemistry technician failed to use appropriate self-checking work practices when verifying the sample results.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Diesel Generator Heat Exchanger Tube Plugging Calculation The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly calculate the tube plugging limit for the emergency diesel generator intercooler, jacket water, and lube oil cooler heat exchangers. The calculation determined that approximately 1/3 of the tubes could be plugged without challenging emergency diesel generator operability under worst case design basis conditions. When corrected, the revised calculation resulted in reducing the allowable number of plugged tubes by approximately 40 percent. The licensee has entered this finding into their corrective action program as Callaway Action Requests 200700063 and 200700096.
The failure to implement appropriate design controls for safety-related tube plugging calculations was a performance deficiency. This finding is more than minor because it affected the mitigating system cornerstone objective (Design Control) to ensure the reliability and capability of the equipment needed to mitigate initiating events. In addition, the
 
finding was more that minor because, if left uncorrected, it could result in a more significant safety concern.
Specifically, if the heat exchanger tubes were plugged to the limit the heat exchangers may be inoperable under certain design basis conditions (i.e., higher essential service water temperatures). Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Essential Service Water Cooling Tower Design Basis Information into Specifications and Procedures.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly translate design requirements into procedures and instructions. Specifically, the cooling tower sizing calculation specified that a flow rate of 15,000 gallons per minute was necessary to meet design basis accident needs but flow balance procedures only required a flow rate of 11,724 gallons per minute. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700218.
The team determined that the failure to properly translate design information (essential service water flow rate through the cooling tower) into specifications and procedures was a performance deficiency. This finding was more than minor because it affected the mitigating system cornerstone objective (Procedure Quality Attribute) to ensure the reliability and capability of the equipment needed to mitigate initiating events. Further, if left uncorrected, it could lead to a more significant issue. Specifically, information from the calculation could be used in other design documents and operability determinations. Over-predicting cooling tower capability could mask other operational issues. Using the Manual Chapter 0609, Phase 1 screening worksheet, the team determined that the finding had very low safety significance (Green) because the finding was a design deficiency confirmed not to result in loss of operability in accordance with Part 9900 Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Initiate an Operability Evaluation for Water Hammer Concerns.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Procedures, for the failure to follow Callaway Plant procedure requirements associated with operability determinations. Specifically, engineers had identified that a water hammer was causing two residual heat removal system relief valves to fail and that the water hammer would likely recur in certain situations. The engineers failed to take the procedurally required actions to initiate a formal operability determination to evaluate the potential impact to the residual heat removal system pressure boundary. The licensee has entered this finding into their corrective action program as Callaway Action Request 200609805.
The failure to follow a Callaway Plant procedure was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Equipment Performance Attribute) of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This finding had a crosscutting aspect in the area of problem identification and resolution (corrective action program component), in that engineers failed to performed the necessary proceduralized corrective actions to ensure that operability was properly evaluated.
Inspection Report# : 2006009 (pdf)
Significance:        Dec 31, 2006
 
Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify a Degraded Fire Barrier On November 21, 2006, the inspectors identified a noncited violation of Technical Specification 5.4.1.d, Fire Protection Program, after AmerenUE failed to identify and correct a degraded auxiliary building fire door. The inspectors identified that the latching mechanism on Fire Door 15031 would not engage because the double door had not been pinned. Failure of the door to latch resulted in a reduction in fire confinement capability. The door was required to provide a 3-hour fire barrier. The licensee had several prior opportunities to identify the degraded fire door. Security and operations personnel passed through the door several times each shift. The inspectors previously identified that the latch on Fire Door 15031 was degraded. Following the previous finding, AmerenUE implemented actions to increase the sensitivity of plant personnel to degraded fire doors. These actions were not effective to ensure that licensee personnel would recognize and enter the degraded fire door into the Corrective Action Program.
This finding is greater than minor because the degraded fire barrier affected the mitigating systems cornerstone external factors attribute objective to prevent undesirable consequences due to fire. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined this finding is in the fire confinement category and that the barrier was moderately degraded because the door latch was not functional. This finding is of very low safety significance because the exposed fire area contained no potential damage targets that are unique from those in the exposing fire area. The inspectors concluded that this finding has a problem identification and resolution crosscutting aspect associated with the corrective action program component because the licensee did not implement the corrective action program with a low threshold to identify the degraded door. The licensee entered this issue into the Corrective Action Program as Callaway Action Request CAR 20060962.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Categorize a Maintenance Preventable Functional Failure The inspectors identified a noncited violation of 10 CFR 50.65(a)(2) after AmerenUE failed to categorize the failure of motor-operated valve auxiliary contacts as a maintenance preventable functional failure and to monitor the component as required by 10 CFR 50.65(a)(1). On May 22, 2006, safety injection system motor-operated Valve EMHV8814A failed to open during surveillance testing due to stuck auxiliary contacts. On June 29, 2006, the Callaway maintenance rule expert panel concluded the failure was not a maintenance preventable functional failure.
The inspectors reviewed the maintenance history of station motor-operated valves and determined eighteen previous auxiliary contact failures had occurred since 2002. Also, AmerenUE had initiated a modification to compensate for motor-operated valve electrical cubicle obsolescence and corrective action to address auxiliary contact failures. The inspectors determined that the June 29, 2006, expert panel incorrectly concluded that the auxiliary contact failures were not maintenance preventable. AmerenUE failed to perform an evaluation as required by 10 CFR 50.65(a)(1). On November 16, 2006 the expert panel reevaluated the failure of Valve EMHV8814A and five other auxiliary contact failures and concluded the failures were maintenance preventable functional failures and placed the auxiliary contacts system in 10 CFR 50.65(a)(1).
This finding is greater than minor because the failure of the expert panel to perform adequate evaluations would become a more significant safety concern if left uncorrected. This issue is similar to Example 7.b provided in Manual Chapter 0612, Appendix E. The inspectors analyzed this finding using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet. The inspectors determined this finding is of very low safety significance because, this finding is not a design or qualification deficiency, did not result in loss of safety function of a single train for greater than the allowed Technical Specification outage time and was not related to a seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the expert panel did not thoroughly or adequately evaluate the failure of the valve to address the causes and extent of condition. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200609603.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Adequately Manage Increased Risk During a Maintenance Activity On September 26, 2006, the inspectors identified a noncited violation of 10 CFR 50.65(a)(4) after AmerenUE failed to adequately manage the risk associated with maintenance on the turbine-driven auxiliary feedwater pump. AmerenUE removed the turbine-driven auxiliary feedwater pump from service for planned maintenance. The licensee determined this activity increased plant risk into the next higher risk configuration (Yellow). Procedure APA-ZZ-00315, Configuration Risk Management Program, required AmerenUE to take actions to protect redundant/diverse safety systems and components. Procedure APA-ZZ-00315 also stated that, if work could result in a risk-significant configuration or loss of system functions, consider use of physical barriers, such as ropes and/or signs to protect redundant/diverse components. AmerenUE did not take adequate protective actions or use physical barriers on the redundant Train B motor-driven auxiliary feedwater pump. Plant workers passing through the motor-driven auxiliary feedwater pump room inadvertently rendered the pump inoperable by disabling the room cooler. The licensee determined that disabling the room cooler increased plant risk into the next higher risk configuration (Orange).
This finding is greater than minor because the licensee failed to implement prescribed significant compensatory measures during planned maintenance activity. This finding is similar to Example 7.g. provided in Manual Chapter 0612, Appendix E, because the auxiliary feedwater system key safety function was degraded. The inspectors used Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 2, Assessment of Risk Management Actions, to analyze this finding. The inspectors calculated an incremental core damage probability of 6.8 x 10-8 for the event, based a one-hour risk exposure duration and an increase of core damage probability from 1.8 x 10-4 to 7.1 x 10-4 after the Train B motor-driven auxiliary feedwater pump inadvertently rendered the pump inoperable. The inspectors determined the finding is of very low safety significance because incremental core damage probability 6.8 x 10-8 was less than 1.0 x 10-6. This finding has a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to appropriately plan work activities by incorporating risk insights and compensatory actions. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 20070284.
Inspection Report# : 2006005 (pdf)
Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Prior NRC Approval Before Removing Technical Specifications Limiting Condition for Operations On October 6, 2006, the inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 after AmerenUE failed to obtain prior NRC approval before removing the steam generator blowdown valve Limiting Condition for Operations requirement from the facility Technical Specifications. Part 50.36 of Title 10 of the Code of Federal Regulations, Technical Specifications, required AmerenUE to establish a Limiting Condition for Operations for components that are required to mitigate a design basis accident. The Callaway Plant accident analysis required the steam generator blowdown valves close to mitigate the steam line break accident and to ensure the auxiliary feedwater system safety function. AmerenUE met this requirement by including the blowdown valves in Technical Specification 3.6.3, Containment Isolation Valves, as referenced in FSAR Table 16.6-1, Containment Isolation Valves. On May 10, 2006, AmerenUE implemented FSAR Change Notice 02-012 which removed the blowdown valves from Table 16.6-1. This change removed the blowdown valves from within the scope the Technical Specifications Limiting Condition for Operations. The 50.59 safety evaluation supporting Change Notice 02-012 failed to identify that removal of the blowdown valves involved a change to the plant Technical Specifications and required prior NRC approval.
This issue involved traditional enforcement because AmerenUE did not receive prior NRC approval before changing the facility Technical Specifications. The inspectors evaluated this issue using Manual Chapter 0612, Appendix B.
This issue is more than minor because the mitigating systems cornerstone attribute of equipment performance, reliability, and capability is impacted based on removal of the blowdown valve out-of-service time limits from the Technical Specifications. The inspectors used Manual Chapter 0609, Significance Determination Process, Phase 1, to analyze the safety significance of the violation. The inspectors concluded that the violation is of very low safety significance because the issue was not a design or qualification deficiency confirmed to result in loss of operability, did not represent a loss of system safety function or an actual loss of safety function of one or more non-Technical Specification risk-significant equipment trains, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The issue has a problem identification and resolution crosscutting aspect associated with the corrective action program because the licensee's safety evaluation did not thoroughly evaluate the
 
change such that the resolutions address causes and extent of conditions, as necessary. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200608902.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of an Operator Workaround Resulted in an Inoperable Safety Injection Accumulator The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after the licensee failed to adequately evaluate and promptly correct an operator workaround that resulted in the loss of nitrogen pressure on a safety injection accumulator. On December 3, 2006, Accumulator D was rendered inoperable due to low pressure. The low pressure condition occurred as plant operators attempted to add nitrogen to the accumulator. Plant operator efforts to work around degraded containment isolation and pressure relief valves during the filling operation resulted in an inoperable accumulator. The accumulator pressure had dropped below the minimum allowed Technical Specification pressure of 602 psig.
This issue is greater than minor because this finding is associated with the reactor safety mitigating systems cornerstone attribute of equipment performance and affects the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the inspectors determined that this finding is of very low significance because, although the condition did involve the loss of operability, it did not result in a loss of system safety train or function, and did not involve a seismic, flooding or severe weather event. This finding, which involved an inadequate evaluation of an operator workaround, has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because AmerenUE did not thoroughly evaluate problems such that resolutions addressed the causes and extent of conditions, as necessary. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200700286.
Inspection Report# : 2006005 (pdf)
Significance:        Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Callaway Action Request The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to initiate Callaway Action Requests for conditions adverse to quality that affected the reliability of mitigating systems.
Specifically, on August 17, 2005, and on May 30, 2006, the licensee discovered a high point air trap in the Train A safety injection discharge piping and decreasing water level in Steam Generators A and D; however, the licensee failed to enter these conditions adverse to quality into their corrective action program. The water in the main steam line contributed to a water hammer and the void had the potential to impact operability of the safety injection system.
The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609812.
The performance deficiency involved the failure to initiate corrective action documents for identified conditions adverse to quality, as required. This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the reliability and availability of systems that respond to initiating events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone, was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. The finding has cross-cutting aspects related to problem identification and resolution, in that, personnel did not identify issues at a low threshold and in a timely manner commensurate with their safety significance.
Inspection Report# : 2006012 (pdf)
Significance:        Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify conditions adverse to quality
 
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, and the corrective action program because licensee personnel failed to recognize and to identify two separate examples as conditions adverse to quality. Specifically, on April 13, 2006, and on October 17, 2006, licensee personnel did not identify blocked containment cooler tubes and a dirty emergency diesel generator turbocharger air intake filter, respectively, as conditions adverse to quality. Failure to recognize these conditions as degraded and identify them as conditions adverse to quality, delayed the immediate evaluation of operability and implementation of corrective actions. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609813.
The performance deficiency involved the failure to promptly identify and correct conditions adverse to quality. The inappropriate classification of Callaway Action Requests 200602989 and 200608806 as Action Notice Callaway Action Requests delayed and prevented actions required by the corrective action program. This finding is greater than minor because a later evaluation by the licensee determined that safety related equipment had been adversely affected.
[This deficiency is similar to Manual Chapter 0612, Appendix E, Example 4.a.] Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. The finding has cross-cutting aspects related to problem identification and resolution, in that, personnel did not identify issues at a low threshold and in a timely manner commensurate with their safety significance.
Inspection Report# : 2006012 (pdf)
Significance:      Nov 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to effectively implement actions to prevent recurrence A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, resulted from the failure to correct, and preclude repetition of (evaluate extent of condition), a significant condition adverse to quality related to identification of high spots in horizontal safety injection system discharge piping. Specifically, the licensee failed to identify all high spots in the susceptible discharge piping in February 2005; consequently, a modification did not prevent recurrence of voids collecting in high spots. The licensee entered the deficiency into their corrective action program as Callaway Action Request 200608644.
The performance deficiency involved the failure to effectively evaluate all susceptible points in the Train A safety injection discharge piping. This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the availability of systems that respond to initiating events. The failure of the design change affected the reliability of the safety injection system. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a cross-cutting aspect related to problem identification and resolution, in that, the licensee did not thoroughly evaluate the voiding problems such that the resolutions addressed the extent of condition.
Inspection Report# : 2006012 (pdf)
Significance:      Nov 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to promptly correct a condition adverse to quality.
A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI resulted after operations personnel failed to implement corrective actions. Specifically, the licensee failed to modify Procedure OSP-AL-V0003, Auxiliary Feedwater Pump Discharge Check Valve (ALV0054) Closure Test, to ensure that upstream piping would be vented prior to performing the test to prevent overpressurizing the turbine-driven auxiliary feedwater pump suction pipe. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200509277.
The performance deficiency involved the failure to change a procedure as recommended in a corrective action to
 
prevent recurrence. This finding associated with failure to implement corrective action is greater than minor because, if left uncorrected, the finding would become a more significant safety concern. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with resources because the licensee did not ensure complete, accurate, up-to-date procedures were available to plant operators.
Inspection Report# : 2006012 (pdf)
Significance: N/A Oct 13, 2006 Identified By: NRC Item Type: FIN Finding Supplemental inspection following a white mitigating systems performance index heat removal system performance indicator.
The U.S. Nuclear Regulatory Commission performed this supplemental inspection to assess the licensees evaluation associated with a performance indicator (Mitigating Systems Performance Index Heat Removal System) that became White with the initial implementation of the Mitigating Systems Performance Index performance indicators during the second quarter of 2006. The primary reason for this performance indicator being characterized as White was system reliability for the auxiliary feedwater system. The licensee performed a comprehensive evaluation that identified three primary root causes for the degraded reliability of the auxiliary feedwater system: poor implementation of maintenance programs to improve quality; a lack of training for maintenance personnel; and poor coordination of personnel and resources. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee, in general, adequately determined the root and contributing causes of the White performance indicator and established appropriate corrective actions. In addition, the licensee conducted an extent of cause review, which included a performance assessment of the remaining mitigating systems.
Inspection Report# : 2006013 (pdf)
Barrier Integrity Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action for Refueling Water Storage Tank Vortexing Concerns The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action) for the failure to take adequate corrective actions following the identification of a condition adverse to quality. Specifically, the licensee had identified, in part, that a safety-related refueling water storage tank sizing calculation had failed to consider vortexing at the tank suction inlet piping. This phenomena can cause air entrainment in pumps, which can lead to pump failure. The corrective measures were inadequate because engineers inappropriately used the margin associated with instrument uncertainty as if it were available design margin. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700224.
The team determined that the failure to take effective corrective measures to address a condition adverse to quality (failure to address vortexing in the refueling water storage tank sizing calculation) was a performance deficiency. The finding was more than minor because it affected the barrier integrity cornerstone objective (design control attribute) to provide reasonable assurance that physical design barriers (including the containment) protect the public from radio nuclide releases caused by accidents or events. The finding had crosscutting aspects in the area of problem identification and resolution (Operating Experience Attribute), in that the licensee had failed to adequately address the industry operating experience.
Inspection Report# : 2006009 (pdf)
Significance:        Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation
 
Inadequate operability determination of a degraded main steam isolation valve The team identified a noncited violation of Technical Specification 3.7.2, after operations personnel failed to enter and implement required Technical Specification 3.7.2 actions. Specifically, the licensee had performed an inadequate operability determination related to a degraded main steam isolation valve that resulted in exceeding the allowed Technical Specifications out-of-service time between December 29 and 31, 2004. On October 19, 2006, the NRC determined that the licensee should have declared the main steam isolation valve and its actuation channel inoperable after removing one of two hydraulic actuators from service. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609233.
The performance deficiency involved the failure to perform an adequate operability evaluation of degraded plant equipment. As a result, the licensee failed to comply with the Technical Specifications. This finding is greater than minor because the configuration control attribute of the barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events is affected. The team used the At Power Significance Determination Process, of Manual Chapter 0609. The team concluded that a Phase 2 analysis was required because this finding affects both the fuel and containment barriers.
The team performed a Phase 2 analysis using the Risk-Informed Inspection Notebook for Callaway Nuclear Generating Station Unit 1, Revision 2. The team assumed that (1) one of two actuator trains was unavailable on one main steam isolation valve for less than 3 days and (2) the degraded actuator did not reduce the remaining main steam isolation valve mitigation capability credit to less than full mitigation credit. Based on the results of the Phase 2 analysis, this finding is determined to have very low safety significance. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly and correctly evaluate the operability of the degraded main steam isolation valve.
Inspection Report# : 2006012 (pdf)
Emergency Preparedness Significance:        Sep 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Program Failure to Ensure Emergency Action Level Entered when Meeting the Defined Limit for Hazardous Atmosphere The inspectors identified a Green noncited violation of 10 CFR 50.54(q) for a failure to adequately implement the emergency plan. The licensee failed to declare an ALERT when conditions existed that met Emergency Action Level 3J, Hazards Affecting Plant Safety. AmerenUE placed this issue in the corrective action program as Callaway Action Request 200607835.
This finding is greater than minor because this finding is associated with the reactor safety emergency preparedness cornerstone attribute of emergency response organization performance and affects the cornerstone objective of the licensee protecting public health and safety during a radiological emergency. The inspectors used Manual Chapter 0609, Significance Determination Process, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, because the licensee misunderstood the emergency action level, but otherwise adequately implemented the emergency plan. The inspectors concluded this finding is of very low safety significance because the performance deficiency is related to the inability to implement one emergency action level at the ALERT level, which is a risk significant planning standard problem but not a risk significant planning standard function failure or a risk significant planning standard degraded function. This finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not provide training to the emergency response organization that clearly communicated the basis for decisions associated with the language changes made to Emergency Action Level 3J.
Inspection Report# : 2006004 (pdf)
Occupational Radiation Safety
 
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Nov 03, 2006 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed 230 Callaway Action Requests, several job orders, engineering evaluations, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities. The team concluded that, generally, the licensee effectively identified, evaluated and prioritized, and implemented effective corrective actions for conditions adverse to quality. However, the team identified that additional effort is needed in all three areas. The team identified some instances of failure to initiate corrective action documents and numerous examples of failure to appropriately classify deficiencies as conditions adverse to quality.
The team determined that quality and documentation for operability assessments has not improved significantly over the course of the evaluation period. Further, on occasion personnel were not self-critical as reflected by poor operational decision making. Two examples of findings reflect the condition of the corrective action problem evaluation activities in the mid portion of the assessment period. The team remained concerned that a lack of understanding of the detailed design and licensing basis continued to be evident in problem resolution. The team concluded that the licensee, generally, implemented timely, effective corrective actions, although some examples indicate continuing weakness in this area.
The team determined that the licensee had increased efforts to evaluate existing industry operating experience for relevance to the facility, and had entered identified items in the corrective action program; however, the team identified some examples that contributed to plant events.
The extensive performance improvement plan developed to address the substantive cross-cutting issue in human performance has addressed daily worker practice issues very well, although recent events occurred that indicate challenges remain. The increased management involvement in the corrective action program and in daily activities assisted in the improved performance. The team determined that licensee audits and assessments became more detailed, probing and self-critical with better assessments at the end of the assessment period. The licensee used benchmarking of industry best practices and third party evaluations that improved the corrective action program during this assessment period. While some of the changes were too recent to evaluate, the team concluded that improvements in the significant root cause process, Corrective Action Review Board graded approach, and scope and timing of corrective actions had improved.
On the basis of formal and informal interviews conducted during this inspection, the team determined that employees will raise issues to their supervision, use the corrective action program, and if necessary, bring concerns to the employee concerns program. The team concluded that the licensee established an acceptable and improving safety-conscious work environment. However, some indication exists that additional effort is needed to encourage the free flow of information to ensure safety issues are resolved promptly.
Inspection Report# : 2006012 (pdf)
Last modified : August 24, 2007
 
Callaway 3Q/2007 Plant Inspection Findings Initiating Events Significance:        Sep 22, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Maintenance Instructions Affecting the Letdown Backpressure Control Valve.
A self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified after the licensee failed to follow reassembly procedures for the letdown system backpressure control valve. In April 2007, during reassembly of letdown pressure control Valve BGPCV0131, Callaway maintenance personnel failed to install an alignment cage spacer. On September 7, 2007, a failed pressure transmitter combined with malfunctioning Valve BGPCV0131 caused upstream letdown relief Valve BG8117 to lift, diverting water into the pressurizer relief tank at a rate of 119 gpm until operators isolated letdown to stop the leakage.
This finding is greater than minor because, similar to Example 5b provided in Manual Chapter 0612, Appendix E, the licensees failure to follow assembly procedures resulted in Valve BGPCV0131 being returned to service with a missing part. This finding, involving reactor coolant system letdown, affected the initiating events cornerstone equipment performance attribute and affected the objective to limit the likelihood of those events that upset plant stability and challenged critical safety functions during power operations. The inspectors used the Manual Chapter 0609, "Significant Determination Process," Phase 1 worksheet to analyze this finding. The inspectors determined this finding is of very low safety significance because it did not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and did not affect any mitigating systems. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to follow established procedures (H.4(b)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200708233.
Inspection Report# : 2007004 (pdf)
Significance:        Mar 24, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Management of an Operator Workaround Resulted in Unplanned Loss of Voume Control Tank Inventory The inspectors identified a finding after volume control tank inventory was inadvertently diverted from the reactor coolant system due to inadequate management of an operator workaround. On January 19 and March 22, 2007, operators had isolated the volume control tank from the demineralizer during resin transfer operations. However, volume control tank inventory was lost due to leakage past closed demineralizer isolation valves. Degraded Grinnell diaphragm valves have been a longstanding Callaway Plant material condition problem. Plant operations did not track nor effectively work around the degraded demineralizer valves.
This finding is greater than minor because the failure to adequately manage operator workarounds could reasonably be viewed as a precursor to a significant event. Using the Manual Chapter 0609, "Significance Determination Process,"
Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because the condition did not result in the reactor coolant system technical specification leakage limit being exceeded, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or internal/external flood. This finding has a crosscutting aspect in the area of human performance associated with the work control component because AmerenUE did not plan work activities to support long-term equipment reliability by limiting operator workarounds (H.3(b)). The licensee entered this finding into their corrective action program as Callaway Action Request 200700517.
Inspection Report# : 2007002 (pdf)
 
Mitigating Systems Significance:        Sep 22, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Promptly Correct a Condition Adverse to Quality for Train B Motor-driven Auxiliary Feedwater Pump A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, was identified after the licensee allowed the Train B motor-driven auxiliary feedwater pump to be returned to service even though maintenance personnel could not meet the coupling shaft separation tolerance during a maintenance activity on April 12, 2007. Engineering personnel approved deviating from the coupling shaft separation tolerance without considering the impact on the motor thrust bearing. On July 4, 2007, motor disassembly revealed that there was damage to the thrust bearing caused by the inadequate shaft separation distance.
This finding is greater than minor because, similar to Example 5b provided in Manual Chapter 0612, Appendix E, the licensees failure to address the impact of plant changes allowed the component to be returned to service prior to correcting the problem. This finding was associated with the mitigating systems cornerstone equipment performance attribute and affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used the Manual Chapter 0609, "Significant Determination Process," Phase 1 worksheet to analyze this finding. The inspectors determined this finding is of very low safety significance because it is not a design or qualification deficiency confirmed to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment; did not result in loss-of-safety function of a single train for greater than the Technical Specification allowed outage time; and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because engineering personnel did not thoroughly evaluate the apparent problem with the coupling (P.1 (c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200708752.
Inspection Report# : 2007004 (pdf)
Significance:        Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Evaluate the Design Basis for an Ultimate Heat Sink Workaround The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after AmerenUE failed to implement effective corrective actions to correct discrepancies in the ultimate heat sink design basis. The system design basis required the ultimate heat sink automated temperature controller to align the cooling tower only when outside temperatures were above 80 degrees Fahrenheit. AmerenUE allowed manual operation of the system when temperatures were above 47 degrees Fahrenheit. The engineering staff and later the quality assurance staff independently identified that the design basis operating requirements had not been adequately evaluated. The inspectors identified that the corrective actions assigned had been closed out as complete without problem resolution and that the ultimate heat sink cooling towers were operated on April 3, 2007, when outside conditions were below 29 degrees Fahrenheit. The uncontrolled workaround resulted in AmerenUE subjecting the cooling tower fill material and fan to freezing conditions.
This finding is greater than minor because it is associated with the mitigating systems cornerstone equipment performance attribute and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it affected the mitigating systems cornerstone, which was both a performance and design deficiency that did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolution would address causes and extent of conditions, as necessary (P.1(c)). This issue was entered into the
 
licensee's corrective action program as Callaway Action Request 200703584.
Inspection Report# : 2007003 (pdf)
Significance:      Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Essential Service Water Pipe Wall Thinning The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after AmerenUEs past corrective actions were inadequate to identify and correct essential service water piping degradation due to corrosion. AmerenUE identified that nondestructive examinations were required to determine the extent of condition of microbiological influenced corrosion on the 30-inch and 8-inch essential service water piping.
On May 3, 2007, operability determinations used to support Refueling Outage 15 restart stated that 100 percent of the low flow area accessible piping would be tested using nondestructive examination. On May 26, 2007, microbiological influenced corrosion caused a new through-wall leak in the control building low flow, accessible piping. The licensees extent of condition review was not adequate to identify the corroded pipe prior to the through-wall leak.
This finding, associated with failure to implement corrective action, is greater than minor because, if left uncorrected, this finding would become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it affected the mitigating systems cornerstone, was both a performance and design deficiency that did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolution would address causes and extent of conditions, as necessary (P.1(c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200705489.
Inspection Report# : 2007003 (pdf)
Significance:      Mar 24, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inoperable Auxiliary Feedwater Pump due to an Inadequae Surveillance Procedure A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an inadequate surveillance procedure resulted in the inadvertent defeat of the Train B turbine-driven auxiliary feedwater pump automatic start feature and an unplanned actuation of a cross-train control room ventilation isolation. On February 12, 2007, plant instrumentation and control technicians were performing a control room ventilation response time test.
The procedure required the operator to block a high radiation test signal. The operator was unable to locate the block switch. A control room supervisor authorized a change to the procedure, which resulted in an incorrect block switch being used. The control room supervisor failed to verify correct block switch identification prior to authorizing the surveillance procedure change.
This finding is greater than minor because the failure to use an adequate surveillance procedure is associated with the mitigating systems cornerstone attribute of procedure quality and affects the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because it was not a design or qualification deficiency, did not result in loss-of-safety function of a single train for greater than the technical specifications allowed outage time, and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the control room supervisor did not thoroughly evaluate the apparent procedure problem before approving the change (P.1 (c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200701336.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 24, 2007 Identified By: NRC Item Type: NCV NonCited Violation
 
Inadequate Corrective Actions to Preserve Essential Service Water System Material Condition The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria XVI, after past corrective actions were inadequate to preclude recurrence of essential service water piping degradation due to corrosion. On March 14 and 23, 2007, plant personnel identified through-wall leaks in the Train B large bore essential service water piping. Plant operators declared the essential service water train inoperable and implemented elevated plant risk and required implementation of risk management actions in both cases. Plant technicians performed non-destructive examinations on about 10 percent of the accessible large bore piping. Technicians identified 93 indications of less than minimum pipe wall thickness. The licensee concluded the pipe degradation resulted from microbiologically influenced corrosion. Poor material condition of the essential service water system has been a longstanding problem at the Callaway Plant. On March 23, 2005, plant personnel identified an essential service water through-wall leak in large bore piping, which required a technical specification required shutdown and on January 25, 2006, plant operators declared Train B of the essential service water system inoperable due to a through-wall pipe leak. These conditions were identified as significant conditions adverse to quality in the licensees corrective action program. The licensee's extent of condition review and corrective actions following the March 23, 2005, and January 25, 2006, occurrences were not adequate to prevent further examples of degraded essential service water piping from microbiologically influenced corrosion.
This finding is greater than minor because it is associated with the reactor safety mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, this finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because AmerenUE did not fully evaluate essential service water corrosion issues to ensure that the resolutions adequately addressed the causes and extent of condition needed to ensure nuclear safety (P.1(c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200702724.
Inspection Report# : 2007002 (pdf)
Significance:      Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Refueling Water Storage Tank Vent Sizing Calculation The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for an inadequate refeuling water storage tank vent sizing calculation. The calculation assumed that only one low head safety injection pump would operate when it should have assumed that all six emergency core cooling and containment spray pumps would take suction from the tank. When corrected, the revised calculation resulted in reducing the allowable vent blockage area from approximately 68 percent to 30 percent. In response to the teams concerns, the licensee inspected the vent and found a small mesh screen on the vents exterior, which reduced the available design margin to approximately 5 percent. Subsequently, the licensee performed a new finite element analysis to demonstrate that sufficient margin existed to account for screen blockage scenarios, such as freezing rain. The licensee has entered this finding into their corrective action program as Callaway Action Requests 200610359 and 200700115.
The failure to meet design control requirements associated with the refeuling water storage tank vent design was a performance deficiency. This finding is more than minor because it affected the mitigating system cornerstone objective (design control attribute) to ensure the reliability and capability of the equipment needed to mitigate initiating events. The finding also affected the barrier integrity cornerstone objective (design control attribute) of providing physical design barriers, such as containment, to protect the public from radio nuclide releases caused by accidents or events. The team used the Manual Chapter 0609, Significance Determination Process Phase 1 screening worksheet and determined that the finding required a Phase 2 significance determination because it impacted two different cornerstones (mitigating systems and barrier integrity). The team performed a Phase 2 significance determination and determined that the finding was of very low safety significance. Only the large break loss-of-coolant accident sequence was affected. In addition, the safety injection and containment spray systems remained available.
Inspection Report# : 2006009 (pdf)
 
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Diesel Generator Fuel Oil Verification The team identified a noncited violation of Technical Specifications Surveillance Requirement 3.8.3.3 for the failure to verify that fuel oil testing results were within the specified limits. Consequently, fuel oil that was transferred to the Train A storage tank in October 2005 was out of specification for cetane and no actions were taken to evaluate or otherwise address the concern until identified by the NRC. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700100.
The failure to follow plant technical specifications and properly verify that the cetane level of new fuel oil was within the limits of the Diesel Fuel Oil Testing Program was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (human performance attribute) of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This finding had a crosscutting aspect in the area of human performance (work practices attribute), in that the chemistry technician failed to use appropriate self-checking work practices when verifying the sample results H.4(a)).
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Diesel Generator Heat Exchanger Tube Plugging Calculation The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly calculate the tube plugging limit for the emergency diesel generator intercooler, jacket water, and lube oil cooler heat exchangers. The calculation determined that approximately 1/3 of the tubes could be plugged without challenging emergency diesel generator operability under worst case design basis conditions. When corrected, the revised calculation resulted in reducing the allowable number of plugged tubes by approximately 40 percent. The licensee has entered this finding into their corrective action program as Callaway Action Requests 200700063 and 200700096.
The failure to implement appropriate design controls for safety-related tube plugging calculations was a performance deficiency. This finding is more than minor because it affected the mitigating system cornerstone objective (Design Control) to ensure the reliability and capability of the equipment needed to mitigate initiating events. In addition, the finding was more that minor because, if left uncorrected, it could result in a more significant safety concern.
Specifically, if the heat exchanger tubes were plugged to the limit the heat exchangers may be inoperable under certain design basis conditions (i.e., higher essential service water temperatures). Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Essential Service Water Cooling Tower Design Basis Information into Specifications and Procedures.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly translate design requirements into procedures and instructions. Specifically, the cooling tower sizing calculation specified that a flow rate of 15,000 gallons per minute was necessary to meet design basis accident needs but flow balance procedures only required a flow rate of 11,724 gallons per minute. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700218.
 
The team determined that the failure to properly translate design information (essential service water flow rate through the cooling tower) into specifications and procedures was a performance deficiency. This finding was more than minor because it affected the mitigating system cornerstone objective (Procedure Quality Attribute) to ensure the reliability and capability of the equipment needed to mitigate initiating events. Further, if left uncorrected, it could lead to a more significant issue. Specifically, information from the calculation could be used in other design documents and operability determinations. Over-predicting cooling tower capability could mask other operational issues. Using the Manual Chapter 0609, Phase 1 screening worksheet, the team determined that the finding had very low safety significance (Green) because the finding was a design deficiency confirmed not to result in loss of operability in accordance with Part 9900 Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Initiate an Operability Evaluation for Water Hammer Concerns.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Procedures, for the failure to follow Callaway Plant procedure requirements associated with operability determinations. Specifically, engineers had identified that a water hammer was causing two residual heat removal system relief valves to fail and that the water hammer would likely recur in certain situations. The engineers failed to take the procedurally required actions to initiate a formal operability determination to evaluate the potential impact to the residual heat removal system pressure boundary. The licensee has entered this finding into their corrective action program as Callaway Action Request 200609805.
The failure to follow a Callaway Plant procedure was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Equipment Performance Attribute) of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This finding had a crosscutting aspect in the area of problem identification and resolution (corrective action program component), in that engineers failed to performed the necessary proceduralized corrective actions to ensure that operability was properly evaluated P.1.(c)).
Inspection Report# : 2006009 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify a Degraded Fire Barrier On November 21, 2006, the inspectors identified a noncited violation of Technical Specification 5.4.1.d, Fire Protection Program, after AmerenUE failed to identify and correct a degraded auxiliary building fire door. The inspectors identified that the latching mechanism on Fire Door 15031 would not engage because the double door had not been pinned. Failure of the door to latch resulted in a reduction in fire confinement capability. The door was required to provide a 3-hour fire barrier. The licensee had several prior opportunities to identify the degraded fire door. Security and operations personnel passed through the door several times each shift. The inspectors previously identified that the latch on Fire Door 15031 was degraded. Following the previous finding, AmerenUE implemented actions to increase the sensitivity of plant personnel to degraded fire doors. These actions were not effective to ensure that licensee personnel would recognize and enter the degraded fire door into the Corrective Action Program.
This finding is greater than minor because the degraded fire barrier affected the mitigating systems cornerstone external factors attribute objective to prevent undesirable consequences due to fire. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined this finding is in the fire confinement category and that the barrier was moderately degraded because the door latch was not functional. This finding is of very low safety significance because the exposed fire area contained no potential damage targets that are unique from those in the exposing fire area. The inspectors concluded that this finding has a problem identification
 
and resolution crosscutting aspect associated with the corrective action program component because the licensee did not implement the corrective action program with a low threshold to identify the degraded door. The licensee entered this issue into the Corrective Action Program as Callaway Action Request CAR 20060962.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Categorize a Maintenance Preventable Functional Failure The inspectors identified a noncited violation of 10 CFR 50.65(a)(2) after AmerenUE failed to categorize the failure of motor-operated valve auxiliary contacts as a maintenance preventable functional failure and to monitor the component as required by 10 CFR 50.65(a)(1). On May 22, 2006, safety injection system motor-operated Valve EMHV8814A failed to open during surveillance testing due to stuck auxiliary contacts. On June 29, 2006, the Callaway maintenance rule expert panel concluded the failure was not a maintenance preventable functional failure.
The inspectors reviewed the maintenance history of station motor-operated valves and determined eighteen previous auxiliary contact failures had occurred since 2002. Also, AmerenUE had initiated a modification to compensate for motor-operated valve electrical cubicle obsolescence and corrective action to address auxiliary contact failures. The inspectors determined that the June 29, 2006, expert panel incorrectly concluded that the auxiliary contact failures were not maintenance preventable. AmerenUE failed to perform an evaluation as required by 10 CFR 50.65(a)(1). On November 16, 2006 the expert panel reevaluated the failure of Valve EMHV8814A and five other auxiliary contact failures and concluded the failures were maintenance preventable functional failures and placed the auxiliary contacts system in 10 CFR 50.65(a)(1).
This finding is greater than minor because the failure of the expert panel to perform adequate evaluations would become a more significant safety concern if left uncorrected. This issue is similar to Example 7.b provided in Manual Chapter 0612, Appendix E. The inspectors analyzed this finding using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet. The inspectors determined this finding is of very low safety significance because, this finding is not a design or qualification deficiency, did not result in loss of safety function of a single train for greater than the allowed Technical Specification outage time and was not related to a seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the expert panel did not thoroughly or adequately evaluate the failure of the valve to address the causes and extent of condition. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200609603.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage Increased Risk During a Maintenance Activity On September 26, 2006, the inspectors identified a noncited violation of 10 CFR 50.65(a)(4) after AmerenUE failed to adequately manage the risk associated with maintenance on the turbine-driven auxiliary feedwater pump. AmerenUE removed the turbine-driven auxiliary feedwater pump from service for planned maintenance. The licensee determined this activity increased plant risk into the next higher risk configuration (Yellow). Procedure APA-ZZ-00315, Configuration Risk Management Program, required AmerenUE to take actions to protect redundant/diverse safety systems and components. Procedure APA-ZZ-00315 also stated that, if work could result in a risk-significant configuration or loss of system functions, consider use of physical barriers, such as ropes and/or signs to protect redundant/diverse components. AmerenUE did not take adequate protective actions or use physical barriers on the redundant Train B motor-driven auxiliary feedwater pump. Plant workers passing through the motor-driven auxiliary feedwater pump room inadvertently rendered the pump inoperable by disabling the room cooler. The licensee determined that disabling the room cooler increased plant risk into the next higher risk configuration (Orange).
This finding is greater than minor because the licensee failed to implement prescribed significant compensatory measures during planned maintenance activity. This finding is similar to Example 7.g. provided in Manual Chapter 0612, Appendix E, because the auxiliary feedwater system key safety function was degraded. The inspectors used Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 2, Assessment of Risk Management Actions, to analyze this finding. The
 
inspectors calculated an incremental core damage probability of 6.8 x 10-8 for the event, based a one-hour risk exposure duration and an increase of core damage probability from 1.8 x 10-4 to 7.1 x 10-4 after the Train B motor-driven auxiliary feedwater pump inadvertently rendered the pump inoperable. The inspectors determined the finding is of very low safety significance because incremental core damage probability 6.8 x 10-8 was less than 1.0 x 10-6. This finding has a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to appropriately plan work activities by incorporating risk insights and compensatory actions. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 20070284.
Inspection Report# : 2006005 (pdf)
Significance: SL-IV Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain Prior NRC Approval Before Removing Technical Specifications Limiting Condition for Operations On October 6, 2006, the inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 after AmerenUE failed to obtain prior NRC approval before removing the steam generator blowdown valve Limiting Condition for Operations requirement from the facility Technical Specifications. Part 50.36 of Title 10 of the Code of Federal Regulations, Technical Specifications, required AmerenUE to establish a Limiting Condition for Operations for components that are required to mitigate a design basis accident. The Callaway Plant accident analysis required the steam generator blowdown valves close to mitigate the steam line break accident and to ensure the auxiliary feedwater system safety function. AmerenUE met this requirement by including the blowdown valves in Technical Specification 3.6.3, Containment Isolation Valves, as referenced in FSAR Table 16.6-1, Containment Isolation Valves. On May 10, 2006, AmerenUE implemented FSAR Change Notice 02-012 which removed the blowdown valves from Table 16.6-1. This change removed the blowdown valves from within the scope the Technical Specifications Limiting Condition for Operations. The 50.59 safety evaluation supporting Change Notice 02-012 failed to identify that removal of the blowdown valves involved a change to the plant Technical Specifications and required prior NRC approval.
This issue involved traditional enforcement because AmerenUE did not receive prior NRC approval before changing the facility Technical Specifications. The inspectors evaluated this issue using Manual Chapter 0612, Appendix B.
This issue is more than minor because the mitigating systems cornerstone attribute of equipment performance, reliability, and capability is impacted based on removal of the blowdown valve out-of-service time limits from the Technical Specifications. The inspectors used Manual Chapter 0609, Significance Determination Process, Phase 1, to analyze the safety significance of the violation. The inspectors concluded that the violation is of very low safety significance because the issue was not a design or qualification deficiency confirmed to result in loss of operability, did not represent a loss of system safety function or an actual loss of safety function of one or more non-Technical Specification risk-significant equipment trains, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The issue has a problem identification and resolution crosscutting aspect associated with the corrective action program because the licensee's safety evaluation did not thoroughly evaluate the change such that the resolutions address causes and extent of conditions, as necessary. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200608902.
Inspection Report# : 2006005 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of an Operator Workaround Resulted in an Inoperable Safety Injection Accumulator The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after the licensee failed to adequately evaluate and promptly correct an operator workaround that resulted in the loss of nitrogen pressure on a safety injection accumulator. On December 3, 2006, Accumulator D was rendered inoperable due to low pressure. The low pressure condition occurred as plant operators attempted to add nitrogen to the accumulator. Plant operator efforts to work around degraded containment isolation and pressure relief valves during the filling operation resulted in an inoperable accumulator. The accumulator pressure had dropped below the minimum allowed Technical Specification pressure of 602 psig.
This issue is greater than minor because this finding is associated with the reactor safety mitigating systems cornerstone attribute of equipment performance and affects the objective to ensure the availability, reliability, and
 
capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the inspectors determined that this finding is of very low significance because, although the condition did involve the loss of operability, it did not result in a loss of system safety train or function, and did not involve a seismic, flooding or severe weather event. This finding, which involved an inadequate evaluation of an operator workaround, has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because AmerenUE did not thoroughly evaluate problems such that resolutions addressed the causes and extent of conditions, as necessary. The licensee entered this issue into the Corrective Action Program as Callaway Action Request 200700286.
Inspection Report# : 2006005 (pdf)
Significance:        Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to initiate Callaway Action Request The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to initiate Callaway Action Requests for conditions adverse to quality that affected the reliability of mitigating systems.
Specifically, on August 17, 2005, and on May 30, 2006, the licensee discovered a high point air trap in the Train A safety injection discharge piping and decreasing water level in Steam Generators A and D; however, the licensee failed to enter these conditions adverse to quality into their corrective action program. The water in the main steam line contributed to a water hammer and the void had the potential to impact operability of the safety injection system.
The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609812.
The performance deficiency involved the failure to initiate corrective action documents for identified conditions adverse to quality, as required. This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the reliability and availability of systems that respond to initiating events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone, was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. The finding has cross-cutting aspects related to problem identification and resolution, in that, personnel did not identify issues at a low threshold and in a timely manner commensurate with their safety significance.
Inspection Report# : 2006012 (pdf)
Significance:        Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify conditions adverse to quality The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, and the corrective action program because licensee personnel failed to recognize and to identify two separate examples as conditions adverse to quality. Specifically, on April 13, 2006, and on October 17, 2006, licensee personnel did not identify blocked containment cooler tubes and a dirty emergency diesel generator turbocharger air intake filter, respectively, as conditions adverse to quality. Failure to recognize these conditions as degraded and identify them as conditions adverse to quality, delayed the immediate evaluation of operability and implementation of corrective actions. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609813.
The performance deficiency involved the failure to promptly identify and correct conditions adverse to quality. The inappropriate classification of Callaway Action Requests 200602989 and 200608806 as Action Notice Callaway Action Requests delayed and prevented actions required by the corrective action program. This finding is greater than minor because a later evaluation by the licensee determined that safety related equipment had been adversely affected.
[This deficiency is similar to Manual Chapter 0612, Appendix E, Example 4.a.] Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. The finding has cross-cutting aspects related to problem identification and resolution, in that, personnel did not identify issues at a low threshold and in a timely manner commensurate with their safety significance.
 
Inspection Report# : 2006012 (pdf)
Significance:        Nov 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to effectively implement actions to prevent recurrence A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, resulted from the failure to correct, and preclude repetition of (evaluate extent of condition), a significant condition adverse to quality related to identification of high spots in horizontal safety injection system discharge piping. Specifically, the licensee failed to identify all high spots in the susceptible discharge piping in February 2005; consequently, a modification did not prevent recurrence of voids collecting in high spots. The licensee entered the deficiency into their corrective action program as Callaway Action Request 200608644.
The performance deficiency involved the failure to effectively evaluate all susceptible points in the Train A safety injection discharge piping. This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective of ensuring the availability of systems that respond to initiating events. The failure of the design change affected the reliability of the safety injection system. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a cross-cutting aspect related to problem identification and resolution, in that, the licensee did not thoroughly evaluate the voiding problems such that the resolutions addressed the extent of condition.
Inspection Report# : 2006012 (pdf)
Significance:        Nov 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to promptly correct a condition adverse to quality.
A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI resulted after operations personnel failed to implement corrective actions. Specifically, the licensee failed to modify Procedure OSP-AL-V0003, Auxiliary Feedwater Pump Discharge Check Valve (ALV0054) Closure Test, to ensure that upstream piping would be vented prior to performing the test to prevent overpressurizing the turbine-driven auxiliary feedwater pump suction pipe. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200509277.
The performance deficiency involved the failure to change a procedure as recommended in a corrective action to prevent recurrence. This finding associated with failure to implement corrective action is greater than minor because, if left uncorrected, the finding would become a more significant safety concern. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design or qualification deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with resources because the licensee did not ensure complete, accurate, up-to-date procedures were available to plant operators.
Inspection Report# : 2006012 (pdf)
Significance: N/A Oct 13, 2006 Identified By: NRC Item Type: FIN Finding Supplemental inspection following a white mitigating systems performance index heat removal system performance indicator.
The U.S. Nuclear Regulatory Commission performed this supplemental inspection to assess the licensees evaluation associated with a performance indicator (Mitigating Systems Performance Index Heat Removal System) that became White with the initial implementation of the Mitigating Systems Performance Index performance indicators during the second quarter of 2006. The primary reason for this performance indicator being characterized as White was system reliability for the auxiliary feedwater system. The licensee performed a comprehensive evaluation that identified three
 
primary root causes for the degraded reliability of the auxiliary feedwater system: poor implementation of maintenance programs to improve quality; a lack of training for maintenance personnel; and poor coordination of personnel and resources. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee, in general, adequately determined the root and contributing causes of the White performance indicator and established appropriate corrective actions. In addition, the licensee conducted an extent of cause review, which included a performance assessment of the remaining mitigating systems.
Inspection Report# : 2006013 (pdf)
Barrier Integrity Significance:      Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Foreign Material Controls for the Refueling Cavity with Reactor Head Removed The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, after refueling personnel did not maintain procedurally required foreign material exclusion barriers.
AmerenUEs foreign material exclusion procedure specified attaching foreign material exclusion curtains to the plant north end of the reactor head missile shield to ensure no foreign material was introduced into the reactor vessel. On April 19, 2007, the inspectors observed the reactor refueling task and noted that there were no curtains acting as the north refueling cavity boundary.
This finding is greater than minor because, if left uncorrected, introduction of foreign material into the reactor cavity would become a more significant safety concern. The barrier integrity cornerstone human performance attribute is used to ensure foreign material and loose parts do not challenge fuel cladding. The inspectors determined this finding to be of very low safety significance using the significance determination process for at-power reactor situations. The inspectors used the at-power significance determination process because of the concern with foreign material impact on an operating reactor core. This finding is of very low safety significance per Inspection Manual Chapter 0609 because the condition was a fuel barrier issue. This finding had a crosscutting aspect in the area of human performance associated with the resources component because plant operators failed to follow procedures established to prevent the introduction of foreign material into the reactor vessel (H.4(b)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200704169.
Inspection Report# : 2007003 (pdf)
Significance:      Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action for Refueling Water Storage Tank Vortexing Concerns The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action) for the failure to take adequate corrective actions following the identification of a condition adverse to quality. Specifically, the licensee had identified, in part, that a safety-related refueling water storage tank sizing calculation had failed to consider vortexing at the tank suction inlet piping. This phenomena can cause air entrainment in pumps, which can lead to pump failure. The corrective measures were inadequate because engineers inappropriately used the margin associated with instrument uncertainty as if it were available design margin. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700224.
The team determined that the failure to take effective corrective measures to address a condition adverse to quality (failure to address vortexing in the refueling water storage tank sizing calculation) was a performance deficiency. The finding was more than minor because it affected the barrier integrity cornerstone objective (design control attribute) to provide reasonable assurance that physical design barriers (including the containment) protect the public from radio nuclide releases caused by accidents or events. The finding had crosscutting aspects in the area of problem identification and resolution (Operating Experience Attribute), in that the licensee had failed to adequately address the industry operating experience P.2(b)).
Inspection Report# : 2006009 (pdf)
 
Significance:        Nov 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate operability determination of a degraded main steam isolation valve The team identified a noncited violation of Technical Specification 3.7.2, after operations personnel failed to enter and implement required Technical Specification 3.7.2 actions. Specifically, the licensee had performed an inadequate operability determination related to a degraded main steam isolation valve that resulted in exceeding the allowed Technical Specifications out-of-service time between December 29 and 31, 2004. On October 19, 2006, the NRC determined that the licensee should have declared the main steam isolation valve and its actuation channel inoperable after removing one of two hydraulic actuators from service. The licensee entered this deficiency into their corrective action program as Callaway Action Request 200609233.
The performance deficiency involved the failure to perform an adequate operability evaluation of degraded plant equipment. As a result, the licensee failed to comply with the Technical Specifications. This finding is greater than minor because the configuration control attribute of the barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events is affected. The team used the At Power Significance Determination Process, of Manual Chapter 0609. The team concluded that a Phase 2 analysis was required because this finding affects both the fuel and containment barriers.
The team performed a Phase 2 analysis using the Risk-Informed Inspection Notebook for Callaway Nuclear Generating Station Unit 1, Revision 2. The team assumed that (1) one of two actuator trains was unavailable on one main steam isolation valve for less than 3 days and (2) the degraded actuator did not reduce the remaining main steam isolation valve mitigation capability credit to less than full mitigation credit. Based on the results of the Phase 2 analysis, this finding is determined to have very low safety significance. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly and correctly evaluate the operability of the degraded main steam isolation valve.
Inspection Report# : 2006012 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Nov 03, 2006 Identified By: NRC
 
Item Type: FIN Finding Identification and Resolution of Problems The team reviewed 230 Callaway Action Requests, several job orders, engineering evaluations, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities. The team concluded that, generally, the licensee effectively identified, evaluated and prioritized, and implemented effective corrective actions for conditions adverse to quality. However, the team identified that additional effort is needed in all three areas. The team identified some instances of failure to initiate corrective action documents and numerous examples of failure to appropriately classify deficiencies as conditions adverse to quality.
The team determined that quality and documentation for operability assessments has not improved significantly over the course of the evaluation period. Further, on occasion personnel were not self-critical as reflected by poor operational decision making. Two examples of findings reflect the condition of the corrective action problem evaluation activities in the mid portion of the assessment period. The team remained concerned that a lack of understanding of the detailed design and licensing basis continued to be evident in problem resolution. The team concluded that the licensee, generally, implemented timely, effective corrective actions, although some examples indicate continuing weakness in this area.
The team determined that the licensee had increased efforts to evaluate existing industry operating experience for relevance to the facility, and had entered identified items in the corrective action program; however, the team identified some examples that contributed to plant events.
The extensive performance improvement plan developed to address the substantive cross-cutting issue in human performance has addressed daily worker practice issues very well, although recent events occurred that indicate challenges remain. The increased management involvement in the corrective action program and in daily activities assisted in the improved performance. The team determined that licensee audits and assessments became more detailed, probing and self-critical with better assessments at the end of the assessment period. The licensee used benchmarking of industry best practices and third party evaluations that improved the corrective action program during this assessment period. While some of the changes were too recent to evaluate, the team concluded that improvements in the significant root cause process, Corrective Action Review Board graded approach, and scope and timing of corrective actions had improved.
On the basis of formal and informal interviews conducted during this inspection, the team determined that employees will raise issues to their supervision, use the corrective action program, and if necessary, bring concerns to the employee concerns program. The team concluded that the licensee established an acceptable and improving safety-conscious work environment. However, some indication exists that additional effort is needed to encourage the free flow of information to ensure safety issues are resolved promptly.
Inspection Report# : 2006012 (pdf)
Last modified : December 07, 2007
 
Callaway 4Q/2007 Plant Inspection Findings Initiating Events Significance:        Sep 22, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Maintenance Instructions Affecting the Letdown Backpressure Control Valve.
A self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified after the licensee failed to follow reassembly procedures for the letdown system backpressure control valve. In April 2007, during reassembly of letdown pressure control Valve BGPCV0131, Callaway maintenance personnel failed to install an alignment cage spacer. On September 7, 2007, a failed pressure transmitter combined with malfunctioning Valve BGPCV0131 caused upstream letdown relief Valve BG8117 to lift, diverting water into the pressurizer relief tank at a rate of 119 gpm until operators isolated letdown to stop the leakage.
This finding is greater than minor because, similar to Example 5b provided in Manual Chapter 0612, Appendix E, the licensees failure to follow assembly procedures resulted in Valve BGPCV0131 being returned to service with a missing part. This finding, involving reactor coolant system letdown, affected the initiating events cornerstone equipment performance attribute and affected the objective to limit the likelihood of those events that upset plant stability and challenged critical safety functions during power operations. The inspectors used the Manual Chapter 0609, "Significant Determination Process," Phase 1 worksheet to analyze this finding. The inspectors determined this finding is of very low safety significance because it did not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and did not affect any mitigating systems. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to follow established procedures (H.4(b)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200708233.
Inspection Report# : 2007004 (pdf)
Significance:        Mar 24, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Management of an Operator Workaround Resulted in Unplanned Loss of Voume Control Tank Inventory The inspectors identified a finding after volume control tank inventory was inadvertently diverted from the reactor coolant system due to inadequate management of an operator workaround. On January 19 and March 22, 2007, operators had isolated the volume control tank from the demineralizer during resin transfer operations. However, volume control tank inventory was lost due to leakage past closed demineralizer isolation valves. Degraded Grinnell diaphragm valves have been a longstanding Callaway Plant material condition problem. Plant operations did not track nor effectively work around the degraded demineralizer valves.
This finding is greater than minor because the failure to adequately manage operator workarounds could reasonably be viewed as a precursor to a significant event. Using the Manual Chapter 0609, "Significance Determination Process,"
Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because the condition did not result in the reactor coolant system technical specification leakage limit being exceeded, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or internal/external flood. This finding has a crosscutting aspect in the area of human performance associated with the work control component because AmerenUE did not plan work activities to support long-term equipment reliability by limiting operator workarounds (H.3(b)). The licensee entered this finding into their corrective action program as Callaway Action Request 200700517.
Inspection Report# : 2007002 (pdf)
 
Mitigating Systems Significance:        Sep 22, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Promptly Correct a Condition Adverse to Quality for Train B Motor-driven Auxiliary Feedwater Pump A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, was identified after the licensee allowed the Train B motor-driven auxiliary feedwater pump to be returned to service even though maintenance personnel could not meet the coupling shaft separation tolerance during a maintenance activity on April 12, 2007. Engineering personnel approved deviating from the coupling shaft separation tolerance without considering the impact on the motor thrust bearing. On July 4, 2007, motor disassembly revealed that there was damage to the thrust bearing caused by the inadequate shaft separation distance.
This finding is greater than minor because, similar to Example 5b provided in Manual Chapter 0612, Appendix E, the licensees failure to address the impact of plant changes allowed the component to be returned to service prior to correcting the problem. This finding was associated with the mitigating systems cornerstone equipment performance attribute and affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used the Manual Chapter 0609, "Significant Determination Process," Phase 1 worksheet to analyze this finding. The inspectors determined this finding is of very low safety significance because it is not a design or qualification deficiency confirmed to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment; did not result in loss-of-safety function of a single train for greater than the Technical Specification allowed outage time; and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because engineering personnel did not thoroughly evaluate the apparent problem with the coupling (P.1 (c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200708752.
Inspection Report# : 2007004 (pdf)
Significance:        Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Evaluate the Design Basis for an Ultimate Heat Sink Workaround The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after AmerenUE failed to implement effective corrective actions to correct discrepancies in the ultimate heat sink design basis. The system design basis required the ultimate heat sink automated temperature controller to align the cooling tower only when outside temperatures were above 80 degrees Fahrenheit. AmerenUE allowed manual operation of the system when temperatures were above 47 degrees Fahrenheit. The engineering staff and later the quality assurance staff independently identified that the design basis operating requirements had not been adequately evaluated. The inspectors identified that the corrective actions assigned had been closed out as complete without problem resolution and that the ultimate heat sink cooling towers were operated on April 3, 2007, when outside conditions were below 29 degrees Fahrenheit. The uncontrolled workaround resulted in AmerenUE subjecting the cooling tower fill material and fan to freezing conditions.
This finding is greater than minor because it is associated with the mitigating systems cornerstone equipment performance attribute and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it affected the mitigating systems cornerstone, which was both a performance and design deficiency that did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolution would address causes and extent of conditions, as necessary (P.1(c)). This issue was entered into the
 
licensee's corrective action program as Callaway Action Request 200703584.
Inspection Report# : 2007003 (pdf)
Significance:      Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Essential Service Water Pipe Wall Thinning The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after AmerenUEs past corrective actions were inadequate to identify and correct essential service water piping degradation due to corrosion. AmerenUE identified that nondestructive examinations were required to determine the extent of condition of microbiological influenced corrosion on the 30-inch and 8-inch essential service water piping.
On May 3, 2007, operability determinations used to support Refueling Outage 15 restart stated that 100 percent of the low flow area accessible piping would be tested using nondestructive examination. On May 26, 2007, microbiological influenced corrosion caused a new through-wall leak in the control building low flow, accessible piping. The licensees extent of condition review was not adequate to identify the corroded pipe prior to the through-wall leak.
This finding, associated with failure to implement corrective action, is greater than minor because, if left uncorrected, this finding would become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it affected the mitigating systems cornerstone, was both a performance and design deficiency that did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolution would address causes and extent of conditions, as necessary (P.1(c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200705489.
Inspection Report# : 2007003 (pdf)
Significance:      Mar 24, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inoperable Auxiliary Feedwater Pump due to an Inadequae Surveillance Procedure A self-revealing noncited violation of Technical Specification 5.4.1.a, "Procedures," was identified after an inadequate surveillance procedure resulted in the inadvertent defeat of the Train B turbine-driven auxiliary feedwater pump automatic start feature and an unplanned actuation of a cross-train control room ventilation isolation. On February 12, 2007, plant instrumentation and control technicians were performing a control room ventilation response time test.
The procedure required the operator to block a high radiation test signal. The operator was unable to locate the block switch. A control room supervisor authorized a change to the procedure, which resulted in an incorrect block switch being used. The control room supervisor failed to verify correct block switch identification prior to authorizing the surveillance procedure change.
This finding is greater than minor because the failure to use an adequate surveillance procedure is associated with the mitigating systems cornerstone attribute of procedure quality and affects the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that this finding is only of very low significance because it was not a design or qualification deficiency, did not result in loss-of-safety function of a single train for greater than the technical specifications allowed outage time, and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the control room supervisor did not thoroughly evaluate the apparent procedure problem before approving the change (P.1 (c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200701336.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 24, 2007 Identified By: NRC Item Type: NCV NonCited Violation
 
Inadequate Corrective Actions to Preserve Essential Service Water System Material Condition The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria XVI, after past corrective actions were inadequate to preclude recurrence of essential service water piping degradation due to corrosion. On March 14 and 23, 2007, plant personnel identified through-wall leaks in the Train B large bore essential service water piping. Plant operators declared the essential service water train inoperable and implemented elevated plant risk and required implementation of risk management actions in both cases. Plant technicians performed non-destructive examinations on about 10 percent of the accessible large bore piping. Technicians identified 93 indications of less than minimum pipe wall thickness. The licensee concluded the pipe degradation resulted from microbiologically influenced corrosion. Poor material condition of the essential service water system has been a longstanding problem at the Callaway Plant. On March 23, 2005, plant personnel identified an essential service water through-wall leak in large bore piping, which required a technical specification required shutdown and on January 25, 2006, plant operators declared Train B of the essential service water system inoperable due to a through-wall pipe leak. These conditions were identified as significant conditions adverse to quality in the licensees corrective action program. The licensee's extent of condition review and corrective actions following the March 23, 2005, and January 25, 2006, occurrences were not adequate to prevent further examples of degraded essential service water piping from microbiologically influenced corrosion.
This finding is greater than minor because it is associated with the reactor safety mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, this finding was determined to have very low safety significance because it only affected the mitigating systems cornerstone and was not a design deficiency, did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because AmerenUE did not fully evaluate essential service water corrosion issues to ensure that the resolutions adequately addressed the causes and extent of condition needed to ensure nuclear safety (P.1(c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200702724.
Inspection Report# : 2007002 (pdf)
Significance:      Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Refueling Water Storage Tank Vent Sizing Calculation The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for an inadequate refeuling water storage tank vent sizing calculation. The calculation assumed that only one low head safety injection pump would operate when it should have assumed that all six emergency core cooling and containment spray pumps would take suction from the tank. When corrected, the revised calculation resulted in reducing the allowable vent blockage area from approximately 68 percent to 30 percent. In response to the teams concerns, the licensee inspected the vent and found a small mesh screen on the vents exterior, which reduced the available design margin to approximately 5 percent. Subsequently, the licensee performed a new finite element analysis to demonstrate that sufficient margin existed to account for screen blockage scenarios, such as freezing rain. The licensee has entered this finding into their corrective action program as Callaway Action Requests 200610359 and 200700115.
The failure to meet design control requirements associated with the refeuling water storage tank vent design was a performance deficiency. This finding is more than minor because it affected the mitigating system cornerstone objective (design control attribute) to ensure the reliability and capability of the equipment needed to mitigate initiating events. The finding also affected the barrier integrity cornerstone objective (design control attribute) of providing physical design barriers, such as containment, to protect the public from radio nuclide releases caused by accidents or events. The team used the Manual Chapter 0609, Significance Determination Process Phase 1 screening worksheet and determined that the finding required a Phase 2 significance determination because it impacted two different cornerstones (mitigating systems and barrier integrity). The team performed a Phase 2 significance determination and determined that the finding was of very low safety significance. Only the large break loss-of-coolant accident sequence was affected. In addition, the safety injection and containment spray systems remained available.
Inspection Report# : 2006009 (pdf)
 
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Diesel Generator Fuel Oil Verification The team identified a noncited violation of Technical Specifications Surveillance Requirement 3.8.3.3 for the failure to verify that fuel oil testing results were within the specified limits. Consequently, fuel oil that was transferred to the Train A storage tank in October 2005 was out of specification for cetane and no actions were taken to evaluate or otherwise address the concern until identified by the NRC. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700100.
The failure to follow plant technical specifications and properly verify that the cetane level of new fuel oil was within the limits of the Diesel Fuel Oil Testing Program was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (human performance attribute) of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This finding had a crosscutting aspect in the area of human performance (work practices attribute), in that the chemistry technician failed to use appropriate self-checking work practices when verifying the sample results H.4(a)).
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Diesel Generator Heat Exchanger Tube Plugging Calculation The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly calculate the tube plugging limit for the emergency diesel generator intercooler, jacket water, and lube oil cooler heat exchangers. The calculation determined that approximately 1/3 of the tubes could be plugged without challenging emergency diesel generator operability under worst case design basis conditions. When corrected, the revised calculation resulted in reducing the allowable number of plugged tubes by approximately 40 percent. The licensee has entered this finding into their corrective action program as Callaway Action Requests 200700063 and 200700096.
The failure to implement appropriate design controls for safety-related tube plugging calculations was a performance deficiency. This finding is more than minor because it affected the mitigating system cornerstone objective (Design Control) to ensure the reliability and capability of the equipment needed to mitigate initiating events. In addition, the finding was more that minor because, if left uncorrected, it could result in a more significant safety concern.
Specifically, if the heat exchanger tubes were plugged to the limit the heat exchangers may be inoperable under certain design basis conditions (i.e., higher essential service water temperatures). Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2006009 (pdf)
Significance:        Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Translate Essential Service Water Cooling Tower Design Basis Information into Specifications and Procedures.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly translate design requirements into procedures and instructions. Specifically, the cooling tower sizing calculation specified that a flow rate of 15,000 gallons per minute was necessary to meet design basis accident needs but flow balance procedures only required a flow rate of 11,724 gallons per minute. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700218.
 
The team determined that the failure to properly translate design information (essential service water flow rate through the cooling tower) into specifications and procedures was a performance deficiency. This finding was more than minor because it affected the mitigating system cornerstone objective (Procedure Quality Attribute) to ensure the reliability and capability of the equipment needed to mitigate initiating events. Further, if left uncorrected, it could lead to a more significant issue. Specifically, information from the calculation could be used in other design documents and operability determinations. Over-predicting cooling tower capability could mask other operational issues. Using the Manual Chapter 0609, Phase 1 screening worksheet, the team determined that the finding had very low safety significance (Green) because the finding was a design deficiency confirmed not to result in loss of operability in accordance with Part 9900 Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2006009 (pdf)
Significance:      Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Initiate an Operability Evaluation for Water Hammer Concerns.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Procedures, for the failure to follow Callaway Plant procedure requirements associated with operability determinations. Specifically, engineers had identified that a water hammer was causing two residual heat removal system relief valves to fail and that the water hammer would likely recur in certain situations. The engineers failed to take the procedurally required actions to initiate a formal operability determination to evaluate the potential impact to the residual heat removal system pressure boundary. The licensee has entered this finding into their corrective action program as Callaway Action Request 200609805.
The failure to follow a Callaway Plant procedure was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone objective (Equipment Performance Attribute) of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This finding had a crosscutting aspect in the area of problem identification and resolution (corrective action program component), in that engineers failed to performed the necessary proceduralized corrective actions to ensure that operability was properly evaluated P.1.(c)).
Inspection Report# : 2006009 (pdf)
Barrier Integrity Significance:      Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Foreign Material Controls for the Refueling Cavity with Reactor Head Removed The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, after refueling personnel did not maintain procedurally required foreign material exclusion barriers.
AmerenUEs foreign material exclusion procedure specified attaching foreign material exclusion curtains to the plant north end of the reactor head missile shield to ensure no foreign material was introduced into the reactor vessel. On April 19, 2007, the inspectors observed the reactor refueling task and noted that there were no curtains acting as the north refueling cavity boundary.
This finding is greater than minor because, if left uncorrected, introduction of foreign material into the reactor cavity would become a more significant safety concern. The barrier integrity cornerstone human performance attribute is used to ensure foreign material and loose parts do not challenge fuel cladding. The inspectors determined this finding to be of very low safety significance using the significance determination process for at-power reactor situations. The
 
inspectors used the at-power significance determination process because of the concern with foreign material impact on an operating reactor core. This finding is of very low safety significance per Inspection Manual Chapter 0609 because the condition was a fuel barrier issue. This finding had a crosscutting aspect in the area of human performance associated with the resources component because plant operators failed to follow procedures established to prevent the introduction of foreign material into the reactor vessel (H.4(b)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200704169.
Inspection Report# : 2007003 (pdf)
Significance:      Jan 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action for Refueling Water Storage Tank Vortexing Concerns The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI (Corrective Action) for the failure to take adequate corrective actions following the identification of a condition adverse to quality. Specifically, the licensee had identified, in part, that a safety-related refueling water storage tank sizing calculation had failed to consider vortexing at the tank suction inlet piping. This phenomena can cause air entrainment in pumps, which can lead to pump failure. The corrective measures were inadequate because engineers inappropriately used the margin associated with instrument uncertainty as if it were available design margin. The licensee has entered this finding into their corrective action program as Callaway Action Request 200700224.
The team determined that the failure to take effective corrective measures to address a condition adverse to quality (failure to address vortexing in the refueling water storage tank sizing calculation) was a performance deficiency. The finding was more than minor because it affected the barrier integrity cornerstone objective (design control attribute) to provide reasonable assurance that physical design barriers (including the containment) protect the public from radio nuclide releases caused by accidents or events. The finding had crosscutting aspects in the area of problem identification and resolution (Operating Experience Attribute), in that the licensee had failed to adequately address the industry operating experience P.2(b)).
Inspection Report# : 2006009 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : February 04, 2008
 
Callaway 1Q/2008 Plant Inspection Findings Initiating Events Significance:      Sep 22, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Maintenance Instructions Affecting the Letdown Backpressure Control Valve.
A self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified after the licensee failed to follow reassembly procedures for the letdown system backpressure control valve. In April 2007, during reassembly of letdown pressure control Valve BGPCV0131, Callaway maintenance personnel failed to install an alignment cage spacer. On September 7, 2007, a failed pressure transmitter combined with malfunctioning Valve BGPCV0131 caused upstream letdown relief Valve BG8117 to lift, diverting water into the pressurizer relief tank at a rate of 119 gpm until operators isolated letdown to stop the leakage.
This finding is greater than minor because, similar to Example 5b provided in Manual Chapter 0612, Appendix E, the licensees failure to follow assembly procedures resulted in Valve BGPCV0131 being returned to service with a missing part. This finding, involving reactor coolant system letdown, affected the initiating events cornerstone equipment performance attribute and affected the objective to limit the likelihood of those events that upset plant stability and challenged critical safety functions during power operations. The inspectors used the Manual Chapter 0609, "Significant Determination Process," Phase 1 worksheet to analyze this finding. The inspectors determined this finding is of very low safety significance because it did not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and did not affect any mitigating systems. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to follow established procedures (H.4(b)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200708233.
Inspection Report# : 2007004 (pdf)
Mitigating Systems Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Needed Test Conditions to Satisfy Technical Specification Surveillance Requirement 3.8.1.18 The inspectors identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, after AmerenUE confirmed that the load shedding emergency load sequencing test could not demonstrate that component cooling water pump breakers would perform satisfactorily in service. On November 19, 2007, AmerenUE determined that quantitative data did not exist to support that component cooling water pump breakers would be capable of closing at Step 1 (5 seconds) of the load shedding emergency load sequence. Technical Specification Surveillance Requirement 3.8.1.18, testing of the emergency load sequencing, required the licensee to verify that load blocks are actuated within +/-10 percent of the specified start time.
This finding, failure to correctly test 4 kV essential bus loading, is more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective to ensure availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it was not a design or qualification deficiency, did not represent a loss of system safety function, did not represent a loss of safety function of a single train for
 
greater than its Technical Specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding was evaluated as not having a crosscutting aspect because it was not reflective of current licensee performance.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage Increased Risk During a Maintenance Activity The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) after AmerenUE operating personnel failed to implement prescribed risk management actions associated with maintenance on the Train B emergency diesel generator. NRC inspectors performed a walkdown of the risk management actions prescribed and noted the omission of the measures to protect the turbine-driven auxiliary feedwater pump. AmerenUEs review determined that operators failed to follow work instructions to post the protective measure.
This finding is greater than minor because it was related to maintenance risk management, the overall plant risk assessed was greater than 1.0 E-6 and the licensee failed to implement some prescribed significant compensatory measures. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 2, Assessment of Risk Management Actions, the inspectors determined this finding to be of very low safety significance because other risk management actions were taken. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because operating personnel did not follow instructions to implement the licensees prescribed risk management actions.
Inspection Report# : 2007005 (pdf)
Significance:        Sep 22, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Promptly Correct a Condition Adverse to Quality for Train B Motor-driven Auxiliary Feedwater Pump A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, was identified after the licensee allowed the Train B motor-driven auxiliary feedwater pump to be returned to service even though maintenance personnel could not meet the coupling shaft separation tolerance during a maintenance activity on April 12, 2007. Engineering personnel approved deviating from the coupling shaft separation tolerance without considering the impact on the motor thrust bearing. On July 4, 2007, motor disassembly revealed that there was damage to the thrust bearing caused by the inadequate shaft separation distance.
This finding is greater than minor because, similar to Example 5b provided in Manual Chapter 0612, Appendix E, the licensees failure to address the impact of plant changes allowed the component to be returned to service prior to correcting the problem. This finding was associated with the mitigating systems cornerstone equipment performance attribute and affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used the Manual Chapter 0609, "Significant Determination Process," Phase 1 worksheet to analyze this finding. The inspectors determined this finding is of very low safety significance because it is not a design or qualification deficiency confirmed to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment; did not result in loss-of-safety function of a single train for greater than the Technical Specification allowed outage time; and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because engineering personnel did not thoroughly evaluate the apparent problem with the coupling (P.1 (c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200708752.
Inspection Report# : 2007004 (pdf)
Significance:        Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Evaluate the Design Basis for an Ultimate Heat Sink Workaround
 
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after AmerenUE failed to implement effective corrective actions to correct discrepancies in the ultimate heat sink design basis. The system design basis required the ultimate heat sink automated temperature controller to align the cooling tower only when outside temperatures were above 80 degrees Fahrenheit. AmerenUE allowed manual operation of the system when temperatures were above 47 degrees Fahrenheit. The engineering staff and later the quality assurance staff independently identified that the design basis operating requirements had not been adequately evaluated. The inspectors identified that the corrective actions assigned had been closed out as complete without problem resolution and that the ultimate heat sink cooling towers were operated on April 3, 2007, when outside conditions were below 29 degrees Fahrenheit. The uncontrolled workaround resulted in AmerenUE subjecting the cooling tower fill material and fan to freezing conditions.
This finding is greater than minor because it is associated with the mitigating systems cornerstone equipment performance attribute and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it affected the mitigating systems cornerstone, which was both a performance and design deficiency that did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolution would address causes and extent of conditions, as necessary (P.1(c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200703584.
Inspection Report# : 2007003 (pdf)
Significance:      Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Essential Service Water Pipe Wall Thinning The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after AmerenUEs past corrective actions were inadequate to identify and correct essential service water piping degradation due to corrosion. AmerenUE identified that nondestructive examinations were required to determine the extent of condition of microbiological influenced corrosion on the 30-inch and 8-inch essential service water piping.
On May 3, 2007, operability determinations used to support Refueling Outage 15 restart stated that 100 percent of the low flow area accessible piping would be tested using nondestructive examination. On May 26, 2007, microbiological influenced corrosion caused a new through-wall leak in the control building low flow, accessible piping. The licensees extent of condition review was not adequate to identify the corroded pipe prior to the through-wall leak.
This finding, associated with failure to implement corrective action, is greater than minor because, if left uncorrected, this finding would become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it affected the mitigating systems cornerstone, was both a performance and design deficiency that did not represent a loss of a safety function, and did not affect seismic, flooding or severe weather initiating events. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolution would address causes and extent of conditions, as necessary (P.1(c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200705489.
Inspection Report# : 2007003 (pdf)
Barrier Integrity Significance:      Mar 14, 2008 Identified By: NRC Item Type: NCV NonCited Violation Nonconservative Technical Specification for Battery Inter-cell Connection Resistances
 
Inspection Report# : 2008006 (pdf)
Significance:      Jun 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Foreign Material Controls for the Refueling Cavity with Reactor Head Removed The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, after refueling personnel did not maintain procedurally required foreign material exclusion barriers.
AmerenUEs foreign material exclusion procedure specified attaching foreign material exclusion curtains to the plant north end of the reactor head missile shield to ensure no foreign material was introduced into the reactor vessel. On April 19, 2007, the inspectors observed the reactor refueling task and noted that there were no curtains acting as the north refueling cavity boundary.
This finding is greater than minor because, if left uncorrected, introduction of foreign material into the reactor cavity would become a more significant safety concern. The barrier integrity cornerstone human performance attribute is used to ensure foreign material and loose parts do not challenge fuel cladding. The inspectors determined this finding to be of very low safety significance using the significance determination process for at-power reactor situations. The inspectors used the at-power significance determination process because of the concern with foreign material impact on an operating reactor core. This finding is of very low safety significance per Inspection Manual Chapter 0609 because the condition was a fuel barrier issue. This finding had a crosscutting aspect in the area of human performance associated with the resources component because plant operators failed to follow procedures established to prevent the introduction of foreign material into the reactor vessel (H.4(b)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200704169.
Inspection Report# : 2007003 (pdf)
Emergency Preparedness Significance:      Aug 10, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee Practices Allow Protective Action Recommendations for Areas Where Protective Action Guides are not Exceeded.
The inspectors identified a noncited violation of 10 CFR 50.54(q), 50.47(b)(10), and 10 CFR Part 50, Appendix E, IV (B), for practices that require licensee protective action recommendations to be made for areas offsite that are not affected by the radiological release, contrary to federal guidance. Programmatic expectations (including on-the-job training) to recommend offsite protective actions for the public in areas where dose assessment has not identified that protective action guides are exceeded is a performance deficiency.
This finding is more than minor because it is not similar to the examples of Manual Chapter 0612, Appendix E, and has the potential to impact public safety. This finding is of very low safety significance because it is a failure to comply with NRC requirements, is associated with Emergency Preparedness Planning Standard 50.47(b)(10), is associated with a risk significant planning standard as defined in Manual Chapter 0609, Appendix B, and is not a risk significant planning standard functional failure or risk significant planning standard degraded function because appropriate licensee protective action recommendations in accordance with federal guidance would be issued for all areas of the emergency planning zone where Protective Action Guides are identified as exceeded. This finding is a noncited violation of 10 CFR 50.54(q) and 50.47(b)(10). The licensee has entered this issue into their corrective action system as Callaway Action Request 200707375. This finding was evaluated as not having a crosscutting aspect.
Inspection Report# : 2007005 (pdf)
Occupational Radiation Safety
 
Significance:        Oct 05, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Adhere to a Radiation Work Permit Requirement The inspectors reviewed a self-revealing, noncited violation of a Technical Specification 5.4.1.a. required procedure that resulted in the external contamination of a work group with two of the four workers receiving internal contamination. Specifically, a work group alarmed the personnel contamination monitors while exiting the Radiological Control Area. The licensee investigated the event and determined that the workers did not use the faceshields as required by their radiation work permit and that the radiation protection technician failed to recognize that the workers did not have them. As corrective action, the licensee developed a plant systems job aid for new and supplemental radiation protection technicians, added the event as operating experience to radiation protection and radiation worker training, and implemented disciplinary action.
The failure to adhere to a radiation work permit requirement is a performance deficiency. This finding is more than minor because it is associated with the occupational radiation safety exposure control attribute and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from unnecessary exposure to radiation. The failure to adhere to a radiation work permit requirement lead to workers unintended and additional personnel exposure. The finding was determined to be of very low safety significance because it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a human performance crosscutting component with an aspect of work practices in human error prevention techniques because the workers did not use peer- and self-checking to ensure the radiation work permit required protective equipment was used.
Inspection Report# : 2007005 (pdf)
Significance:        Oct 05, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to notify radiation protection of an electronic dosimeter alarm.
The inspectors reviewed a self-revealing, noncited violation of a Technical Specification 5.4.1.a. required procedure that resulted when mechanical maintenance workers did not report electronic dosimeter alarms when received.
Specifically, during troubleshooting of a waste gas compressor failure, two mechanical maintenance workers received an electronic dosimeter dose rate alarm. The workers exited the room, checked their dosimeters for a dose alarm, determined the noise was due to the compressor and returned to work as no worker had a visible dose alarm. The workers failed to recognize that the electronic dosimeters would not hold and display a dose rate alarm once out of the elevated radiation field. The workers did not notify radiation protection of the electronic dosimeter alarms. When exiting the radiological control area, the electronic dosimeter system alerted the workers to the alarms and barred them from the radiological control area for further entries. As corrective action, the workers were coached on expected dosimeter alarm response and the mechanical maintenance supervisor discussed the event during their group meeting as a learning experience.
The failure to notify radiation protection of an electronic dosimeter alarm is a performance deficiency. This finding is more than minor because it is associated with the occupational radiation safety exposure control attribute and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from unnecessary exposure to radiation. The failure to notify radiation protection in the event of an electronic dosimeter alarm could lead to a workers unintended and additional personnel exposure. The finding was determined to be of very low safety significance because it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a human performance crosscutting component with an aspect of work practices in human error prevention techniques because the workers proceeded in the face of uncertainty or unexpected circumstances when dose rate alarms were received.
Inspection Report# : 2007005 (pdf)
Public Radiation Safety
 
Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : June 05, 2008
 
Callaway 2Q/2008 Plant Inspection Findings Initiating Events Significance:      Sep 22, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Maintenance Instructions Affecting the Letdown Backpressure Control Valve.
A self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified after the licensee failed to follow reassembly procedures for the letdown system backpressure control valve. In April 2007, during reassembly of letdown pressure control Valve BGPCV0131, Callaway maintenance personnel failed to install an alignment cage spacer. On September 7, 2007, a failed pressure transmitter combined with malfunctioning Valve BGPCV0131 caused upstream letdown relief Valve BG8117 to lift, diverting water into the pressurizer relief tank at a rate of 119 gpm until operators isolated letdown to stop the leakage.
This finding is greater than minor because, similar to Example 5b provided in Manual Chapter 0612, Appendix E, the licensees failure to follow assembly procedures resulted in Valve BGPCV0131 being returned to service with a missing part. This finding, involving reactor coolant system letdown, affected the initiating events cornerstone equipment performance attribute and affected the objective to limit the likelihood of those events that upset plant stability and challenged critical safety functions during power operations. The inspectors used the Manual Chapter 0609, "Significant Determination Process," Phase 1 worksheet to analyze this finding. The inspectors determined this finding is of very low safety significance because it did not result in exceeding the Technical Specification limit for identified reactor coolant system leakage and did not affect any mitigating systems. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to follow established procedures (H.4(b)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200708233.
Inspection Report# : 2007004 (pdf)
Mitigating Systems Significance:      Mar 14, 2008 Identified By: NRC Item Type: NCV NonCited Violation Nonconservative Technical Specification for Battery Inter-cell Connection Resistances The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control,"
because the licensee failed to ensure that Technical Specification Surveillance Requirements for the NK11 and NK14 safety related batteries established limits that met the design requirements. Specifically, until questioned by the team the licensee failed to determine the required design value needed to assure plant safety as requested in Callaway Action Request 200706561. The licensee determined that 69 micro ohms should be the actual allowed inter cell voltage limit to meet the design requirements versus an allowed Technical Specification limit of 150 micro ohms.
The performance deficiency associated with this finding involved the failure to ensure that the NK11 and NK14 safety related batteries would remain operable if all the inter cell connections measured 150 micro ohms as allowed by Technical Specification Surveillance Requirements 3.8.4.2 and 3.8.4.5. This finding was greater than minor because it was associated with the Mitigating Systems cornerstone attribute of maintenance and testing and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the finding was determined to have very low safety significance because it was a design deficiency confirmed not to result in loss of operability. The finding had a cross cutting aspect in the area of
 
problem identification and resolution associated with operating experience because the licensee failed to evaluate in a timely manner relevant internal and external operating experience P.2(a) (Section 4OA2.e).
Inspection Report# : 2008006 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Needed Test Conditions to Satisfy Technical Specification Surveillance Requirement 3.8.1.18 The inspectors identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, after AmerenUE confirmed that the load shedding emergency load sequencing test could not demonstrate that component cooling water pump breakers would perform satisfactorily in service. On November 19, 2007, AmerenUE determined that quantitative data did not exist to support that component cooling water pump breakers would be capable of closing at Step 1 (5 seconds) of the load shedding emergency load sequence. Technical Specification Surveillance Requirement 3.8.1.18, testing of the emergency load sequencing, required the licensee to verify that load blocks are actuated within +/-10 percent of the specified start time.
This finding, failure to correctly test 4 kV essential bus loading, is more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective to ensure availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it was not a design or qualification deficiency, did not represent a loss of system safety function, did not represent a loss of safety function of a single train for greater than its Technical Specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding was evaluated as not having a crosscutting aspect because it was not reflective of current licensee performance.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage Increased Risk During a Maintenance Activity The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) after AmerenUE operating personnel failed to implement prescribed risk management actions associated with maintenance on the Train B emergency diesel generator. NRC inspectors performed a walkdown of the risk management actions prescribed and noted the omission of the measures to protect the turbine-driven auxiliary feedwater pump. AmerenUEs review determined that operators failed to follow work instructions to post the p3rotective measure.
This finding is greater than minor because it was related to maintenance risk management, the overall plant risk assessed was greater than 1.0 E-6 and the licensee failed to implement some prescribed significant compensatory measures. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 2, Assessment of Risk Management Actions, the inspectors determined this finding to be of very low safety significance because other risk management actions were taken. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because operating personnel did not follow instructions to implement the licensees prescribed risk management actions. [H.4(b)]
Inspection Report# : 2007005 (pdf)
Significance:        Sep 22, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Promptly Correct a Condition Adverse to Quality for Train B Motor-driven Auxiliary Feedwater Pump A self-revealing Green noncited violation of 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, was identified after the licensee allowed the Train B motor-driven auxiliary feedwater pump to be returned to service even
 
though maintenance personnel could not meet the coupling shaft separation tolerance during a maintenance activity on April 12, 2007. Engineering personnel approved deviating from the coupling shaft separation tolerance without considering the impact on the motor thrust bearing. On July 4, 2007, motor disassembly revealed that there was damage to the thrust bearing caused by the inadequate shaft separation distance.
This finding is greater than minor because, similar to Example 5b provided in Manual Chapter 0612, Appendix E, the licensees failure to address the impact of plant changes allowed the component to be returned to service prior to correcting the problem. This finding was associated with the mitigating systems cornerstone equipment performance attribute and affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used the Manual Chapter 0609, "Significant Determination Process," Phase 1 worksheet to analyze this finding. The inspectors determined this finding is of very low safety significance because it is not a design or qualification deficiency confirmed to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment; did not result in loss-of-safety function of a single train for greater than the Technical Specification allowed outage time; and was not a potentially risk significant seismic, flooding, or severe weather event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because engineering personnel did not thoroughly evaluate the apparent problem with the coupling (P.1 (c)). This issue was entered into the licensee's corrective action program as Callaway Action Request 200708752.
Inspection Report# : 2007004 (pdf)
Barrier Integrity Emergency Preparedness Significance:        Aug 10, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee Practices Allow Protective Action Recommendations for Areas Where Protective Action Guides are not Exceeded.
The inspectors identified a noncited violation of 10 CFR 50.54(q), 50.47(b)(10), and 10 CFR Part 50, Appendix E, IV (B), for practices that require licensee protective action recommendations to be made for areas offsite that are not affected by the radiological release, contrary to federal guidance. Programmatic expectations (including on-the-job training) to recommend offsite protective actions for the public in areas where dose assessment has not identified that protective action guides are exceeded is a performance deficiency.
This finding is more than minor because it is not similar to the examples of Manual Chapter 0612, Appendix E, and has the potential to impact public safety. This finding is of very low safety significance because it is a failure to comply with NRC requirements, is associated with Emergency Preparedness Planning Standard 50.47(b)(10), is associated with a risk significant planning standard as defined in Manual Chapter 0609, Appendix B, and is not a risk significant planning standard functional failure or risk significant planning standard degraded function because appropriate licensee protective action recommendations in accordance with federal guidance would be issued for all areas of the emergency planning zone where Protective Action Guides are identified as exceeded. This finding is a noncited violation of 10 CFR 50.54(q) and 50.47(b)(10). The licensee has entered this issue into their corrective action system as Callaway Action Request 200707375. This finding was evaluated as not having a crosscutting aspect.
Inspection Report# : 2007005 (pdf)
Occupational Radiation Safety Significance:        Oct 05, 2007
 
Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Adhere to a Radiation Work Permit Requirement The inspectors reviewed a self-revealing, noncited violation of a Technical Specification 5.4.1.a. required procedure that resulted in the external contamination of a work group with two of the four workers receiving internal contamination. Specifically, a work group alarmed the personnel contamination monitors while exiting the Radiological Control Area. The licensee investigated the event and determined that the workers did not use the faceshields as required by their radiation work permit and that the radiation protection technician failed to recognize that the workers did not have them. As corrective action, the licensee developed a plant systems job aid for new and supplemental radiation protection technicians, added the event as operating experience to radiation protection and radiation worker training, and implemented disciplinary action.
The failure to adhere to a radiation work permit requirement is a performance deficiency. This finding is more than minor because it is associated with the occupational radiation safety exposure control attribute and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from unnecessary exposure to radiation. The failure to adhere to a radiation work permit requirement lead to workers unintended and additional personnel exposure. The finding was determined to be of very low safety significance because it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a human performance crosscutting component with an aspect of work practices in human error prevention techniques because the workers did not use peer- and self-checking to ensure the radiation work permit required protective equipment was used. [H.4(a)]
Inspection Report# : 2007005 (pdf)
Significance:        Oct 05, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to notify radiation protection of an electronic dosimeter alarm.
The inspectors reviewed a self-revealing, noncited violation of a Technical Specification 5.4.1.a. required procedure that resulted when mechanical maintenance workers did not report electronic dosimeter alarms when received.
Specifically, during troubleshooting of a waste gas compressor failure, two mechanical maintenance workers received an electronic dosimeter dose rate alarm. The workers exited the room, checked their dosimeters for a dose alarm, determined the noise was due to the compressor and returned to work as no worker had a visible dose alarm. The workers failed to recognize that the electronic dosimeters would not hold and display a dose rate alarm once out of the elevated radiation field. The workers did not notify radiation protection of the electronic dosimeter alarms. When exiting the radiological control area, the electronic dosimeter system alerted the workers to the alarms and barred them from the radiological control area for further entries. As corrective action, the workers were coached on expected dosimeter alarm response and the mechanical maintenance supervisor discussed the event during their group meeting as a learning experience.
The failure to notify radiation protection of an electronic dosimeter alarm is a performance deficiency. This finding is more than minor because it is associated with the occupational radiation safety exposure control attribute and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from unnecessary exposure to radiation. The failure to notify radiation protection in the event of an electronic dosimeter alarm could lead to a workers unintended and additional personnel exposure. The finding was determined to be of very low safety significance because it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a human performance crosscutting component with an aspect of work practices in human error prevention techniques because the workers proceeded in the face of uncertainty or unexpected circumstances when dose rate alarms were received. [H.4(a)]
Inspection Report# : 2007005 (pdf)
Public Radiation Safety
 
Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : August 29, 2008
 
Callaway 3Q/2008 Plant Inspection Findings Initiating Events Mitigating Systems Significance: SL-IV Sep 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit a Licensee Event Report for a Condition Prohibitied by the Plant's Technical Specifications The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(1) for a failure to submit a required licensee event report within 60 days after discovery of an event requiring a report. On May 21, 2008, Callaway Plant personnel discovered a 6.6 cubic foot void of air within the safety injection system common suction piping. The voided piping, determined to have existed for over a year, was caused by relief valve maintenance on Valve EM8858A that occurred on May 7, 2007. Callaway Plant licensing staff performed a reportability evaluation and determined that the discovery of the void was not required to be reported to the NRC. The inspectors reviewed the licensees reportability evaluation and associated past operability and determined the event was reportable since a postulated single failure had the potential to disable both emergency core cooling system trains during cold leg recirculation. Since the emergency core cooling system was inoperable from May 7, 2007, until May 21, 2008, the event resulted in an operation or condition which was prohibited by the plants Technical Specifications as well as an event where a single cause or condition caused two independent trains to become inoperable in a single system.
This finding is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. This finding affected the mitigating systems cornerstone. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV, noncited violation. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810199. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate a void discovered in the emergency core cooling system for operability and reportability.
Inspection Report# : 2008004 (pdf)
Significance:        Jun 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Surveillance Procedure Resulted in an Inaperable Emergency Core Cooling System The inspectors identified a noncited violation of Technical Specification 3.5.2, "Emergency Core Cooling Systems," after an inadequate surveillance procedure resulted in the licensee failing to maintain the emergency core cooling system full of water as required per Technical Specification 3.5.2. On May 21, 2008, Callaway Plant engineering discovered that a section of the cold leg recirculation piping, specifically the discharge of the residual heat removal pumps to the safety injection pumps, contained 6.6 cubic feet of air. Callaway monthly surveillance Procedure OSP SA 00003, "Emergency Core Cooling Flow Path Verification and Venting," had a purpose to: "Verify the ECCS is full of water," in accordance with Technical Specification Surveillance Requirement 3.5.2.3. The monthly verification and vent procedure was not comprehensive enough to ensure all the emergency core cooling system was full of water.
This finding is more than minor because it was similar to Example 3e of NRC Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," and met the Not Minor If, criteria because the failure to meet the licensees administrative requirement for allowable void fraction impacted the ability of the Train A safety injection system to function upon initiation of high-pressure recirculation. This finding affected the mitigating systems cornerstone procedure quality attribute. Using the Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding should be evaluated using the Phase 2 process described in Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. As described in Section III, of Appendix A, given that the presolved table did not contain a suitable target or surrogate for this finding, the senior reactor analyst used the risk-informed notebook to evaluate the significance of this finding affecting only high-pressure recirculation as very low risk significance (Green). This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions in decision making and did not adopt a requirement to demonstrate that a single vent valve was sufficient to vent the affected line rather than assuming that an additional installed valve was not necessary to completely fill, vent, and test the line [H.1(b)].
Inspection Report# : 2008003 (pdf)
 
Significance:        Jun 24, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct a Condition Adverse to Quality for Diesel Generator Jacket Water O-Rings A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified after the licensee failed to promptly correct leakage from diesel generator jacket water o-rings. On February 20, 2008, during a normal surveillance run of Emergency Diesel Generator B, Callaway operations personnel identified an approximately 80 drops per minute jacket water leak caused by premature failure of Nitrile type o-rings. Following restoration of Emergency Diesel Generator B, the licensee re-evaluated the preventative maintenance frequency for jacket water o-ring replacement and reduced the replacement frequency from once every three years to once every refueling cycle. Then, on May 28, 2008, during a routine surveillance run of Emergency Diesel Generator A, Callaway operations personnel identified that Emergency Diesel Generator A had a 200 drops per minute jacket water leak. Similar to the condition observed on Emergency Diesel Generator B on February 20, 2008, the source of the leakage was from Nitrile type o-rings within the jacket water system. The o-rings responsible for jacket water leakage were found to be of similar age to those that failed during the February 20, 2008 surveillance but had not been replaced despite the change to the licensee's preventive maintenance frequency.
This finding, failure to implement adequate corrective actions for degraded Nitrile type o-rings in Emergency Diesel Generator A after previously identifying the adverse condition on Emergency Diesel Generator B, is more than minor because, if left uncorrected, degraded diesel generator jacket water o-rings could become a more significant safety concern. This finding affected the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in loss of operability. This finding has a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to plan work activities to support long-term equipment reliability by addressing known degraded conditions in a more reactive than preventative manner [H.3(b)].
Inspection Report# : 2008003 (pdf)
Significance:        Jun 24, 2008 Identified By: NRC Item Type: VIO Violation Failure to Prevent Recurrence of Voids in Emergency Core Cooling System Cold Leg Recirculation Piping The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," because the licensee failed to restore compliance within a reasonable time by establishing measures to prevent void formation in emergency core cooling system suction piping for the Train A safety injection system. On May 21, 2008, Callaway Plant engineering performed ultrasonic inspection of the safety injection system common suction piping Line EM¬023 HCB - 6" and discovered a 6.6 cubic foot voided area. This exceeded the allowable void fraction of 2.1 cubic feet required for operability. This voided piping, determined to have existed for over a year, was caused by relief valve maintenance on Valve EM8858A (May 7, 2007). The maintenance restoration failed to perform a fill and vent to ensure the suction pipe was full of water. The inspectors identified several related examples where the licensee had performed either inadequate operating experience evaluations, inadequate extent of condition reviews, or inadequate procedure corrections.
This finding, failure to restore compliance to prevent recurrence of emergency core cooling system voids was more than minor because it is similar to Example 3e of NRC Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," criteria because the failure impacted the ability of the emergency core cooling system to function upon initiation of high-pressure recirculation. Using the Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding should be evaluated using the Phase 2 process described in Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. As described in Section III, of Appendix A, given that the presolved table did not contain a suitable target or surrogate for this finding, the senior reactor analyst used the risk-informed notebook to evaluate the significance of this finding as very low risk significance (Green). This finding has a crosscutting aspect in the area of problem identification and resolution associated witht the corrective action program component because AmerenUE failed to thoroughly evaluate voiding problems such that the resolutions addressed causes and extent of condition, as necessary [P.1(c)].
Inspection Report# : 2008003 (pdf)
Significance:        Mar 14, 2008 Identified By: NRC Item Type: NCV NonCited Violation Nonconservative Technical Specification for Battery Inter-cell Connection Resistances The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," because the licensee failed to ensure that Technical Specification Surveillance Requirements for the NK11 and NK14 safety related batteries established limits that met the design requirements. Specifically, until questioned by the team the licensee failed to determine the required design value needed to assure plant safety as requested in Callaway Action Request 200706561. The licensee determined that 69 micro ohms should be the actual allowed inter cell voltage limit to meet the design requirements versus an allowed Technical Specification limit of 150 micro ohms.
The performance deficiency associated with this finding involved the failure to ensure that the NK11 and NK14 safety related batteries would remain operable if all the inter cell connections measured 150 micro ohms as allowed by Technical Specification Surveillance Requirements
 
3.8.4.2 and 3.8.4.5. This finding was greater than minor because it was associated with the Mitigating Systems cornerstone attribute of maintenance and testing and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the finding was determined to have very low safety significance because it was a design deficiency confirmed not to result in loss of operability. The finding had a cross cutting aspect in the area of problem identification and resolution associated with operating experience because the licensee failed to evaluate in a timely manner relevant internal and external operating experience P.2(a)
(Section 4OA2.e).
Inspection Report# : 2008006 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Needed Test Conditions to Satisfy Technical Specification Surveillance Requirement 3.8.1.18 The inspectors identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, after AmerenUE confirmed that the load shedding emergency load sequencing test could not demonstrate that component cooling water pump breakers would perform satisfactorily in service. On November 19, 2007, AmerenUE determined that quantitative data did not exist to support that component cooling water pump breakers would be capable of closing at Step 1 (5 seconds) of the load shedding emergency load sequence. Technical Specification Surveillance Requirement 3.8.1.18, testing of the emergency load sequencing, required the licensee to verify that load blocks are actuated within +/-10 percent of the specified start time.
This finding, failure to correctly test 4 kV essential bus loading, is more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective to ensure availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance because it was not a design or qualification deficiency, did not represent a loss of system safety function, did not represent a loss of safety function of a single train for greater than its Technical Specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding was evaluated as not having a crosscutting aspect because it was not reflective of current licensee performance.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Manage Increased Risk During a Maintenance Activity The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) after AmerenUE operating personnel failed to implement prescribed risk management actions associated with maintenance on the Train B emergency diesel generator. NRC inspectors performed a walkdown of the risk management actions prescribed and noted the omission of the measures to protect the turbine-driven auxiliary feedwater pump. AmerenUEs review determined that operators failed to follow work instructions to post the p3rotective measure.
This finding is greater than minor because it was related to maintenance risk management, the overall plant risk assessed was greater than 1.0 E-6 and the licensee failed to implement some prescribed significant compensatory measures. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 2, Assessment of Risk Management Actions, the inspectors determined this finding to be of very low safety significance because other risk management actions were taken. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because operating personnel did not follow instructions to implement the licensees prescribed risk management actions. [H.4(b)]
Inspection Report# : 2007005 (pdf)
Barrier Integrity Significance:        Sep 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Boric Acid Corrosion Control Procedures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform a corrosion evaluation of boric acid leakage from containment spray Valve ENHV0006 in accordance with Procedure EDP ZZ 01004, Boric Acid Corrosion Control Program. On August 29, 2008, the resident inspectors identified an active packing leak on Valve ENHV0006 with impact to carbon steel components on the valve as evident by discolored, brown boron. The leak, which had been active since February 27, 2007, was caused by a stem imperfection that was previously identified on December 5, 2007. The inspectors noted that Valve ENHV0006 did not have a current boric acid corrosion evaluation despite meeting the screening requirements for an evaluation listed in Procedure EDP ZZ 01004, Boric Acid Corrosion Control Program, Section 4.2. Programmatic boric acid control and work control issues
 
were a key contributor to not recognizing the need for an updated boric acid corrosion evaluation.
This finding is more than minor because, if left uncorrected, the failure to analyze the effects of boric acid corrosion on safety related components could become a more significant safety concern. This finding affected the barrier integrity cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined be of very low safety significance because the finding does not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere, does not represent an actual open pathway in the physical integrity of the reactor containment, and does not involve an actual reduction in function of hydrogen ignitors in the reactor containment. This issue was entered into the licensee's corrective action program as Callaway Action Request 200809351. This finding has a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to interdepartmentally coordinate the impact of changes to the work scope for Valve ENHV0006 such that appropriate personnel could perform the necessary evaluations to assure plant performance.
Inspection Report# : 2008004 (pdf)
Significance:        Jun 24, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Ensure the Suitability of the Design of the Containment Air Cooler Control Circuitry A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified after determining that the licensee had not adequately selected and reviewed the suitability of the design of the containment air cooler control circuitry. On March 26, 2008, containment air Cooler A fan shut down when shifted from fast to slow speed. Troubleshooting by the licensee determined that voltage was lost to the control power circuitry when the fast speed thermal overload tripped. Since the overload contacts were wired in series, containment air Cooler A experienced a complete loss of control power rendering it inoperable. The licensee determined the trip to be caused by operation of containment air coolers in fast speed, during a period of higher than normal containment pressure. The licensee analyzed the potential impact of the newly discovered adverse containment cooler design vulnerability against design basis accident scenarios. The licensee determined that a hot zero power main steam line break results in a delayed safety injection signal allowing the fan motor overloads to trip prior to being shed by the load sequencer. The containment air coolers would then experience a complete loss of control power and would not be capable of automatically restarting in slow speed. The analysis revealed that the peak containment pressure limit of 48.1 psig would be preserved. The licensee submitted a Licensee Event Report as required by 10 CFR 50.73 since the inadequate containment air cooler control circuitry resulted in a condition prohibited by the plants Technical Specifications.
This finding, failure to ensure the design of the containment air cooler control circuitry was suitable for all plant conditions, was more than minor because it was associated with the barrier integrity cornerstone attribute of design control and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609 Appendix H, Containment Integrity Significance Determination Process," this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since containment coolers are structures, systems or components that have no impact on large early release frequency. The inspectors determined that this finding does not have a crosscutting aspect associated with it since the performance deficiency was not indicative of current licensee performance.
Inspection Report# : 2008003 (pdf)
Significance:        Jun 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Technical Specificaion Bases Change Process The inspectors identified a noncited violation of Technical Specification 5.4.1.a, Procedures, after Callaway control room operators improperly entered a wrong Technical Specification action statement due to the failure to maintain the Technical Specification Bases current.
On June 17, 2008, during surveillance testing, Valve EMHV8823 failed to indicate fully closed. Since EMHV8823 is an isolation valve for containment Penetration 49, the licensee entered Technical Specification 3.6.3, Containment Isolation Valves, Condition C, with an action to restore the valve to an operable status or isolate the penetration within 72 hours. Approximately 8 hours after valve EMHV8823 had been declared inoperable, Callaway licensing personnel contacted the control room and informed them of an approved Technical Specification Bases change that did not allow Technical Specification 3.6.3 Condition C to be applicable to containment Penetration 49. The Technical Specification Bases change was effective May 1, 2008 but had not been issued to the control room. The licensee determined that the more restrictive Technical Specification 3.6.3, Condition A, should have been entered with an action to isolate the affected penetration within 4 hours. The licensee performed a containment entry following discovery of entry into Technical Specification 3.6.3, Condition A and found that Valve EMHV8823 failed its surveillance due to out of adjustment position indicator limit switches. The valve was verified closed and isolated allowing exit from Technical Specification 3.6.3, Condition A.
This finding, failure to ensure the Technical Specification Bases were maintained current and available to the Callaway control room staff, is more than minor because if left uncorrected, the failure to maintain the Technical Specification Bases current could become a more significant safety concern. This finding was determined to affect the barrier integrity cornerstone. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, this finding is determined to be of very low safety significance since this finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen ignitors in the reactor containment. This finding has a crosscutting aspect in the area of human performance associated with the
 
decision making component because the licensee failed to communicate, in a timely manner, decisions to personnel who have a need to know the information in order to perform work safely [H.1(c)].
Inspection Report# : 2008003 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Oct 05, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Adhere to a Radiation Work Permit Requirement The inspectors reviewed a self-revealing, noncited violation of a Technical Specification 5.4.1.a. required procedure that resulted in the external contamination of a work group with two of the four workers receiving internal contamination. Specifically, a work group alarmed the personnel contamination monitors while exiting the Radiological Control Area. The licensee investigated the event and determined that the workers did not use the faceshields as required by their radiation work permit and that the radiation protection technician failed to recognize that the workers did not have them. As corrective action, the licensee developed a plant systems job aid for new and supplemental radiation protection technicians, added the event as operating experience to radiation protection and radiation worker training, and implemented disciplinary action.
The failure to adhere to a radiation work permit requirement is a performance deficiency. This finding is more than minor because it is associated with the occupational radiation safety exposure control attribute and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from unnecessary exposure to radiation. The failure to adhere to a radiation work permit requirement lead to workers unintended and additional personnel exposure. The finding was determined to be of very low safety significance because it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a human performance crosscutting component with an aspect of work practices in human error prevention techniques because the workers did not use peer- and self-checking to ensure the radiation work permit required protective equipment was used. [H.4(a)]
Inspection Report# : 2007005 (pdf)
Significance:        Oct 05, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to notify radiation protection of an electronic dosimeter alarm.
The inspectors reviewed a self-revealing, noncited violation of a Technical Specification 5.4.1.a. required procedure that resulted when mechanical maintenance workers did not report electronic dosimeter alarms when received. Specifically, during troubleshooting of a waste gas compressor failure, two mechanical maintenance workers received an electronic dosimeter dose rate alarm. The workers exited the room, checked their dosimeters for a dose alarm, determined the noise was due to the compressor and returned to work as no worker had a visible dose alarm. The workers failed to recognize that the electronic dosimeters would not hold and display a dose rate alarm once out of the elevated radiation field. The workers did not notify radiation protection of the electronic dosimeter alarms. When exiting the radiological control area, the electronic dosimeter system alerted the workers to the alarms and barred them from the radiological control area for further entries. As corrective action, the workers were coached on expected dosimeter alarm response and the mechanical maintenance supervisor discussed the event during their group meeting as a learning experience.
The failure to notify radiation protection of an electronic dosimeter alarm is a performance deficiency. This finding is more than minor because it is associated with the occupational radiation safety exposure control attribute and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from unnecessary exposure to radiation. The failure to notify radiation protection in the event of an electronic dosimeter alarm could lead to a workers unintended and additional personnel exposure. The finding was determined to be of very low safety significance because it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a human performance crosscutting component with an aspect of work practices in human error prevention techniques because the workers proceeded in the face of uncertainty or unexpected circumstances when dose rate alarms were received. [H.4(a)]
Inspection Report# : 2007005 (pdf)
 
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : November 26, 2008
 
Callaway 4Q/2008 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to maintain an adequate plant shutdown procedure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after improper isolation of the main steam isolation valves by the Callaway control room operators resulted in a reactor trip signal and auxiliary feedwater actuation on October 11, 2008. Procedure OTG ZZ 00006, "Plant Cooldown Hot Standby to Cold Shutdown," allowed premature main steam isolation valve closures just after entering Mode 4. The operator then decided to reopen main steam isolation Valve A and atmospheric Steam Dump A. This created a significant increase in steam flow from the steam generator which caused the steam generator level to swell up to the P 14 steam generator high level feedwater isolation setpoint. The steam generator levels all decreased to the steam generator narrow range low-low setpoint generating the need for auxiliary feedwater actuation.
This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings," this finding is determined to be of very low safety significance since this finding did not affect the Technical Specification limit for reactor coolant system leakage, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to communicate, in a timely manner, decisions to personnel who have a need to know the information in order to perform work safely.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Failure to evaluate material equivalencies leads to a manual reactor trip The inspectors identified a self-revealing finding for failure of the engineering department to perform a material equivalency evaluation to ensure replacement components do not adversely affect plant operations. On November 11, 2008, Callaway Plant experienced a trip of main feedwater Pump B due to low lube oil pressure. Since the reactor was at greater than 80 percent power, the plant operators inserted a manual reactor trip. Following the reactor trip, maintenance personnel discovered two pieces of o-ring foreign material within main feedwater Pump B bearing oil supply pressure regulating Valve FCV0970. The foreign material was found wrapped around the regulating spring which inhibited valve movement and caused the lube oil low pressure condition. The licensee determined that the ethylene propylene diene M-class type o-ring became pliable when exposed to lube oil and was allowed to fall and be introduced into the system as foreign material. The ethylene propylene diene M-class o-rings had been approved as an equivalent replacement in July 1999 for the vendor recommended Buna-N type o-rings without performing an engineering material equivalency evaluation. Buna-N material is approved for use in petroleum based systems while ethylene propylene diene M-class is not.
This finding is greater than minor because it is associated with the design control attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using Manual Chapter 0609.04, "Phase 1 -
Initial Screening and Characterization of Findings," the finding is determined to be potentially risk significant because it contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not
 
be available. When evaluated per Manual Chapter 0609 Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," and the Callaway Plant Phase 2 pre-solved table item Failure to Reestablish Main Feedwater, the inspectors determined this finding to be of very low safety significance. This issue was entered into the licensee's corrective action program as Callaway Action Request 200811781. This finding was determined to not have a crosscutting aspect because the performance deficiency is not indicative of current licensee performance.
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance:        Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate shutdown risk assessment for maintenance activities in the reactor building.
The inspectors identified a noncited violation of 10 CFR 50.65(a)(4), for failure to adequately assess and manage shutdown risk associated with maintenance activities in the reactor building. Specifically, on October 15, 2008, the inspectors found foreign material exclusion covers installed on the Train B containment recirculation sump. The covers were installed on October 14, 2008, per the direction of the containment coordinator without notification to the control room. The covers were installed to prevent debris from entering the sump. Following discussions with operations personnel, the inspectors found that the Train B containment recirculation sump was inappropriately credited in the licensees shutdown safety assessment. An updated shutdown safety assessment was performed and it was determined that plant risk remained yellow.
This finding is greater than minor because the licensees risk assessment inappropriately credited risk-significant structures, systems and components that were unavailable during maintenance. This finding affected the Mitigating Systems cornerstone. Using Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the finding was found to be of very low safety significance because the licensee maintained two trains of decay heat removal operable and adequate equipment was available to support feed and bleed operations for at least 24 hours. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810540.
This finding had a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to obtain interdisciplinary input on safety-significant or risk-significant decisions. Specifically, the containment coordinator made a decision affecting the availability of the containment recirculation sumps without consulting the control room to determine the impact on plant risk.
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to ensure the suitability of the design of the resideual heat removal Train A pump room cooler The inspectors identified a self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after a trip of the residual heat removal Train A room cooler fan revealed that AmerenUE had not adequately selected and reviewed the suitability of the newly installed fan motor thermal overloads. Additionally, the NRC inspectors identified that the postmaintenance testing prescribed for the modified fan motor breaker did not allow sufficient time to challenge the thermal overload settings. On October 8, 2008, residual heat removal Train A room cooler fan shut down after only 22 minutes of run time. The breaker replacement modification used a calculation originally performed for the initial design of the old breaker which did not account for the cooler fan motor being a 20 horsepower motor nameplated down to a 10 horsepower rating.
This finding is greater than minor because it is similar to Manual Chapter 0612 "Examples of Minor Issues," Example 3j, in that the engineering calculation error resulted in a condition where there was a reasonable doubt on the
 
operability of the component and a significant programmatic deficiency associated with postmaintenance test requirements was identified that could lead to worse errors if uncorrected. The inspectors determined that the finding impacted the Mitigating Systems cornerstone. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than Technical Specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810223. The inspectors determined that this finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action component because the AmerenUE modification for certain motor control center breakers failed to have a low enough threshold to identify fan motor rating and thermal overload setting errors.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to adequately implement plant equipment control tagout procedure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after improper restoration of the essential service water supply to the emergency diesel generator Train A lubricating oil cooler resulted in significant water flow into the emergency diesel room on October 22, 2008. Two restoration evolutions associated with the essential service water and the emergency diesel generator systems had been proceeding in parallel. The reactor operator restoring the emergency diesel generator assumed the essential service water supply was to remain isolated to the emergency diesel generator and thus changed the already approved worker protection assurance Clearance 71899 to leave the oil cooler drain valve open with no tag. Starting the essential service water pump pressurized the drain valve and produced significant water spray flow into the emergency diesel generator room until noticed by a diesel vendor representative about 30 minutes later.
This finding was greater than minor because if left uncorrected the deficiencies could become a more significant safety concern. The finding affected the Mitigating Systems cornerstone. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings," this finding is determined to be of very low safety significance since this finding was not a design or qualification deficiency, did not represent a loss of system or train safety function and did not screen as potentially risk significant due to a flooding initiating event using the criteria on the characterization worksheet. This finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee's pre-job briefing, self- and peer-checking, and proper documentation of activity were inadequate to overcome worker protection assurance clearance process problems and an inexperienced operating supervisor. These less than adequate worker practices resulted in personnel proceeding in the face of uncertainty.
Inspection Report# : 2008005 (pdf)
Significance: SL-IV Sep 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit a Licensee Event Report for a Condition Prohibitied by the Plant's Technical Specifications The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(1) for a failure to submit a required licensee event report within 60 days after discovery of an event requiring a report. On May 21, 2008, Callaway Plant personnel discovered a 6.6 cubic foot void of air within the safety injection system common suction piping. The voided piping, determined to have existed for over a year, was caused by relief valve maintenance on Valve EM8858A that occurred on May 7, 2007. Callaway Plant licensing staff performed a reportability evaluation and determined that the discovery of the void was not required to be reported to the NRC. The inspectors reviewed the licensees reportability evaluation and associated past operability and determined the event was reportable since a postulated single failure had the potential to disable both emergency core cooling system trains during cold leg recirculation. Since the emergency core cooling system was inoperable from May 7, 2007, until May 21, 2008, the event resulted in an operation or condition which was prohibited by the plants Technical Specifications as well as an
 
event where a single cause or condition caused two independent trains to become inoperable in a single system.
This finding is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. This finding affected the mitigating systems cornerstone. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV, noncited violation. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810199. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate a void discovered in the emergency core cooling system for operability and reportability.
Inspection Report# : 2008004 (pdf)
Significance:      Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Safety Related 125 Vdc Station Battery NK11 Inadequate Battery Sizing Calculation The team identified a non-cited violation 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to verify the adequacy of design and for failure to correctly translate the 125 Vdc system design basis into instructions, procedures, and drawings. Specifically, the licensee failed to include momentary loads in the battery sizing calculation, thus reducing the peak load demand voltage during the first minute of an event, an intermediate scenario event, and the last minute of the battery duty cycle. Additionally, the licensees subsequent review determined that the calculation had failed to include three additional momentary loads. The failure to include these loads prevented the licensee from developing a battery duty cycle profile that conforms to the guidance of IEEE 485-1983 and correctly simulates the battery loads following a design basis or station blackout event. The licensee entered this finding into their corrective action program as Callaway Action Request 200808609.
The failure to account for all loads, including momentary loads, in the battery design calculation was a performance deficiency because it prevented the licensee from correctly analyzing available voltage at safety-related components during the battery peak loading periods. The finding was more than minor because it is associated with the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the safety-related battery systems to respond to initiating events and prevent undesirable consequences.
Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because adequate margins had been included in the battery selection and, therefore, the issue was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2008008 (pdf)
Significance:      Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Pipe Break Location for the Condensate Stoarge Tank Supply to Auxiliary Feedwater Pumps The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for not verifying the adequacy of the design. Specifically, an incorrect pipe break location in the analysis for loss of the condensate storage tank feed to the auxiliary feedwater pumps caused the analysis to be non-conservative for the amount of water available to the auxiliary feedwater pumps. This error provided for more water to be available for use by the auxiliary feedwater pumps than would actually be available if the analysis had identified the correct location of the postulated pipe break. The licensee has entered this finding into their corrective action program as Callaway Action Request CAR 200808674.
The failure to meet design control requirements associated with the pipe break analysis with sufficient water to run the auxiliary feedwater pumps prior to switch over to the essential service water system is a performance deficiency. Per
 
Manual Chapter 0612, Appendix E, Section 3, Non-significant Dimensional, Time, Calculation, or Drawing Discrepancies, Example J, this finding is more than minor because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system or component. Using Manual Chapter 0609, Significance Determination Process Phase 1 screening worksheet, the team determined that the finding was of very low safety significance. There was no actual loss of safety function and the new analysis demonstrated that the auxiliary feedwater pumps would have enough water available from the Condensate Storage Tank prior to switchover to the Essential Service Water system to complete their design function.
Inspection Report# : 2008008 (pdf)
Significance:        Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater Turbine Digital Control Panel FC219 The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawing for the failure to follow Procedure APA-ZZ-00500, Appendix 1, Revision 6, Operability Determination.
The evaluation did not include the additional heat loading on equipment in the turbine driven auxiliary feedwater pump room, caused from an active steam leak from the turbine governor end case joint. The licensee had failed to include the additional steam leak heat load in either of the room temperature calculations M-GF-415 or BO -05, which were used in the operability determination. The heat input into the room, due to the steam leak, may have adversely affected the operation of the turbine digital speed control unit. The licensee has entered this finding into their corrective action program as Callaway Action Request 200808777.
The failure to either correct the active steam leak or to account for the leak in their design calculations, is a performance deficiency. Per Inspection Manual Chapter 0612, Appendix E, Section 3, Non-significant Dimensional, Time, Calculation, or Drawing Discrepancies, Example J, this finding is more than minor because the licensee had not resolved the deficiency, resulting in a condition in which there was a reasonable doubt regarding the reliability of the turbine digital speed control unit. Using Inspection Manual Chapter 0609, "Significance Determination Process,"
Phase 1 screening worksheets, the team determined that the finding was of very low safety significance. Since there was no actual loss of safety function and the new analysis demonstrated that the maximum room temperature, including the additional heat load, would not exceed the design limit of digital turbine speed controls unit, the issue was a design deficiency confirmed not to result in loss of operability per NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The finding had crosscutting aspects in the area of human performance (decision making) because the licensee used non-conservative assumptions in decision making and failed to either repair the active steam leak, or to account for it in their design calculations. This activity was indicative of current performance as the steam leak still existed and had not been included in the design calculations until October 2008.
Inspection Report# : 2008008 (pdf)
Significance:        Jun 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Surveillance Procedure Resulted in an Inaperable Emergency Core Cooling System The inspectors identified a noncited violation of Technical Specification 3.5.2, "Emergency Core Cooling Systems,"
after an inadequate surveillance procedure resulted in the licensee failing to maintain the emergency core cooling system full of water as required per Technical Specification 3.5.2. On May 21, 2008, Callaway Plant engineering discovered that a section of the cold leg recirculation piping, specifically the discharge of the residual heat removal pumps to the safety injection pumps, contained 6.6 cubic feet of air. Callaway monthly surveillance Procedure OSP SA 00003, "Emergency Core Cooling Flow Path Verification and Venting," had a purpose to: "Verify the ECCS is full of water," in accordance with Technical Specification Surveillance Requirement 3.5.2.3. The monthly verification and vent procedure was not comprehensive enough to ensure all the emergency core cooling system was full of water.
This finding is more than minor because it was similar to Example 3e of NRC Inspection Manual Chapter 0612,
 
Appendix E, "Examples of Minor Issues," and met the Not Minor If, criteria because the failure to meet the licensees administrative requirement for allowable void fraction impacted the ability of the Train A safety injection system to function upon initiation of high-pressure recirculation. This finding affected the mitigating systems cornerstone procedure quality attribute. Using the Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding should be evaluated using the Phase 2 process described in Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. As described in Section III, of Appendix A, given that the presolved table did not contain a suitable target or surrogate for this finding, the senior reactor analyst used the risk-informed notebook to evaluate the significance of this finding affecting only high-pressure recirculation as very low risk significance (Green). This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions in decision making and did not adopt a requirement to demonstrate that a single vent valve was sufficient to vent the affected line rather than assuming that an additional installed valve was not necessary to completely fill, vent, and test the line [H.1(b)].
Inspection Report# : 2008003 (pdf)
Significance:        Jun 24, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct a Condition Adverse to Quality for Diesel Generator Jacket Water O-Rings A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified after the licensee failed to promptly correct leakage from diesel generator jacket water o-rings. On February 20, 2008, during a normal surveillance run of Emergency Diesel Generator B, Callaway operations personnel identified an approximately 80 drops per minute jacket water leak caused by premature failure of Nitrile type o-rings.
Following restoration of Emergency Diesel Generator B, the licensee re-evaluated the preventative maintenance frequency for jacket water o-ring replacement and reduced the replacement frequency from once every three years to once every refueling cycle. Then, on May 28, 2008, during a routine surveillance run of Emergency Diesel Generator A, Callaway operations personnel identified that Emergency Diesel Generator A had a 200 drops per minute jacket water leak. Similar to the condition observed on Emergency Diesel Generator B on February 20, 2008, the source of the leakage was from Nitrile type o-rings within the jacket water system. The o-rings responsible for jacket water leakage were found to be of similar age to those that failed during the February 20, 2008 surveillance but had not been replaced despite the change to the licensee's preventive maintenance frequency.
This finding, failure to implement adequate corrective actions for degraded Nitrile type o-rings in Emergency Diesel Generator A after previously identifying the adverse condition on Emergency Diesel Generator B, is more than minor because, if left uncorrected, degraded diesel generator jacket water o-rings could become a more significant safety concern. This finding affected the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in loss of operability. This finding has a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to plan work activities to support long-term equipment reliability by addressing known degraded conditions in a more reactive than preventative manner [H.3(b)].
Inspection Report# : 2008003 (pdf)
Significance:        Jun 24, 2008 Identified By: NRC Item Type: VIO Violation Failure to Prevent Recurrence of Voids in Emergency Core Cooling System Cold Leg Recirculation Piping The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," because the licensee failed to restore compliance within a reasonable time by establishing measures to prevent void formation in emergency core cooling system suction piping for the Train A safety injection system. On May 21, 2008, Callaway Plant engineering performed ultrasonic inspection of the safety injection system common suction piping Line EM¬023 HCB - 6" and discovered a 6.6 cubic foot voided area. This exceeded the allowable void fraction of 2.1 cubic feet required for operability. This voided piping, determined to have existed for over a year, was caused by
 
relief valve maintenance on Valve EM8858A (May 7, 2007). The maintenance restoration failed to perform a fill and vent to ensure the suction pipe was full of water. The inspectors identified several related examples where the licensee had performed either inadequate operating experience evaluations, inadequate extent of condition reviews, or inadequate procedure corrections.
This finding, failure to restore compliance to prevent recurrence of emergency core cooling system voids was more than minor because it is similar to Example 3e of NRC Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," criteria because the failure impacted the ability of the emergency core cooling system to function upon initiation of high-pressure recirculation. Using the Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding should be evaluated using the Phase 2 process described in Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. As described in Section III, of Appendix A, given that the presolved table did not contain a suitable target or surrogate for this finding, the senior reactor analyst used the risk-informed notebook to evaluate the significance of this finding as very low risk significance (Green). This finding has a crosscutting aspect in the area of problem identification and resolution associated witht the corrective action program component because AmerenUE failed to thoroughly evaluate voiding problems such that the resolutions addressed causes and extent of condition, as necessary [P.1(c)].
Inspection Report# : 2008003 (pdf)
Inspection Report# : 2008005 (pdf)
Significance:      Mar 14, 2008 Identified By: NRC Item Type: NCV NonCited Violation Nonconservative Technical Specification for Battery Inter-cell Connection Resistances The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control,"
because the licensee failed to ensure that Technical Specification Surveillance Requirements for the NK11 and NK14 safety related batteries established limits that met the design requirements. Specifically, until questioned by the team the licensee failed to determine the required design value needed to assure plant safety as requested in Callaway Action Request 200706561. The licensee determined that 69 micro ohms should be the actual allowed inter cell voltage limit to meet the design requirements versus an allowed Technical Specification limit of 150 micro ohms.
The performance deficiency associated with this finding involved the failure to ensure that the NK11 and NK14 safety related batteries would remain operable if all the inter cell connections measured 150 micro ohms as allowed by Technical Specification Surveillance Requirements 3.8.4.2 and 3.8.4.5. This finding was greater than minor because it was associated with the Mitigating Systems cornerstone attribute of maintenance and testing and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the finding was determined to have very low safety significance because it was a design deficiency confirmed not to result in loss of operability. The finding had a cross cutting aspect in the area of problem identification and resolution associated with operating experience because the licensee failed to evaluate in a timely manner relevant internal and external operating experience P.2(a) (Section 4OA2.e).
Inspection Report# : 2008006 (pdf)
Barrier Integrity Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate maintenance procedure resulted in residual heat removal mechanical seal failure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1a, Procedures, for
 
inadequate procedural guidance that resulted in the failure of the residual heat removal Train A pump mechanical seal.
On October 22, 2008, the licensee discovered a solid stream of water issuing from the residual heat removal Train A pump mechanical seal. The failure occurred because of installation difficulties encountered on October 8, 2008, when the seal sleeve was installed with the seal locking collar engaged. This configuration resulted in increased loading on the seal seating surfaces that resulted in surface chipping and led to seal failure after approximately 48 hours of shutdown cooling operation. Mechanical seal replacement Procedure MPM EJ QP001, Residual Heat Removal Pump Overhaul, did not specify that the seal sleeve needed to be installed prior to installing the seal-locking collar.
Additionally, the installation procedure did not specify any post-installation acceptance criteria to ensure the seal is properly seated. An analysis of the seal failure determined that leakage would not exceed the 2 gallon per minute Technical Specification limit but would exceed the 1 gallon per minute administrative limit for emergency core cooling system leakage outside containment.
This finding is more than minor because it was associated with the Barrier Integrity cornerstone attribute of procedural quality and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process," this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since the 2 gallon per minute limit assumed in the post accident dose calculation was preserved and therefore the degraded condition would have no impact on large early release frequency. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810933. This finding did not have a crosscutting aspect since it was not a performance deficiency indicative of current licensee performance.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to terminate refueling water storage tank recirculation results in inadvertent loss of spent fuel pool inventory The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1a, Procedures, for the failure to close Valve BNV0002 during a fill of the spent fuel pool resulting in approximately 2000 gallons of water being inadvertently transferred from the spent fuel pool to the refueling water storage tank. On November 7, 2008, Procedure OTN EC 00001 was performed to add makeup water to the spent fuel pool. Prior to performing the evolution, operations briefed that the refueling water storage tank was on recirculation and that this alignment needed to be secured prior to performing a fill of the spent fuel pool. Following termination of the refueling water storage tank recirculation lineup and after a fill of the spent fuel pool was initiated, the control room received annunciator RWST Lev HILO. The crew recognized that an inadvertent transfer of spent fuel pool water to the refueling water storage tank was in progress and directed that Valves ECV0076 and BNV0002 be closed. It was later discovered that poor communication between operators on the status of Valve BNV0002 resulted in the refueling water storage tank remaining on recirculation during the fill operation.
This finding is more than minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because it only represents a degradation of the radiological barrier function provided by the spent fuel pool. This issue was entered into the licensee's corrective action program as Callaway Action Request 200811692.
This finding had a crosscutting aspect in the area of human performance associated with the work control component because operations personnel failed to effectively communicate work status to the control room.
Inspection Report# : 2008005 (pdf)
Significance:      Sep 24, 2008 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Implement Boric Acid Corrosion Control Procedures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform a corrosion evaluation of boric acid leakage from containment spray Valve ENHV0006 in accordance with Procedure EDP ZZ 01004, Boric Acid Corrosion Control Program. On August 29, 2008, the resident inspectors identified an active packing leak on Valve ENHV0006 with impact to carbon steel components on the valve as evident by discolored, brown boron. The leak, which had been active since February 27, 2007, was caused by a stem imperfection that was previously identified on December 5, 2007. The inspectors noted that Valve ENHV0006 did not have a current boric acid corrosion evaluation despite meeting the screening requirements for an evaluation listed in Procedure EDP ZZ 01004, Boric Acid Corrosion Control Program, Section 4.2. Programmatic boric acid control and work control issues were a key contributor to not recognizing the need for an updated boric acid corrosion evaluation.
This finding is more than minor because, if left uncorrected, the failure to analyze the effects of boric acid corrosion on safety related components could become a more significant safety concern. This finding affected the barrier integrity cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined be of very low safety significance because the finding does not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere, does not represent an actual open pathway in the physical integrity of the reactor containment, and does not involve an actual reduction in function of hydrogen ignitors in the reactor containment. This issue was entered into the licensee's corrective action program as Callaway Action Request 200809351. This finding has a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to interdepartmentally coordinate the impact of changes to the work scope for Valve ENHV0006 such that appropriate personnel could perform the necessary evaluations to assure plant performance.
Inspection Report# : 2008004 (pdf)
Significance:      Jun 24, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Ensure the Suitability of the Design of the Containment Air Cooler Control Circuitry A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified after determining that the licensee had not adequately selected and reviewed the suitability of the design of the containment air cooler control circuitry. On March 26, 2008, containment air Cooler A fan shut down when shifted from fast to slow speed. Troubleshooting by the licensee determined that voltage was lost to the control power circuitry when the fast speed thermal overload tripped. Since the overload contacts were wired in series, containment air Cooler A experienced a complete loss of control power rendering it inoperable. The licensee determined the trip to be caused by operation of containment air coolers in fast speed, during a period of higher than normal containment pressure. The licensee analyzed the potential impact of the newly discovered adverse containment cooler design vulnerability against design basis accident scenarios. The licensee determined that a hot zero power main steam line break results in a delayed safety injection signal allowing the fan motor overloads to trip prior to being shed by the load sequencer. The containment air coolers would then experience a complete loss of control power and would not be capable of automatically restarting in slow speed. The analysis revealed that the peak containment pressure limit of 48.1 psig would be preserved. The licensee submitted a Licensee Event Report as required by 10 CFR 50.73 since the inadequate containment air cooler control circuitry resulted in a condition prohibited by the plants Technical Specifications.
This finding, failure to ensure the design of the containment air cooler control circuitry was suitable for all plant conditions, was more than minor because it was associated with the barrier integrity cornerstone attribute of design control and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609 Appendix H, Containment Integrity Significance Determination Process," this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since containment coolers are structures, systems or components that have no impact on large early
 
release frequency. The inspectors determined that this finding does not have a crosscutting aspect associated with it since the performance deficiency was not indicative of current licensee performance.
Inspection Report# : 2008003 (pdf)
Significance:      Jun 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Technical Specificaion Bases Change Process The inspectors identified a noncited violation of Technical Specification 5.4.1.a, Procedures, after Callaway control room operators improperly entered a wrong Technical Specification action statement due to the failure to maintain the Technical Specification Bases current. On June 17, 2008, during surveillance testing, Valve EMHV8823 failed to indicate fully closed. Since EMHV8823 is an isolation valve for containment Penetration 49, the licensee entered Technical Specification 3.6.3, Containment Isolation Valves, Condition C, with an action to restore the valve to an operable status or isolate the penetration within 72 hours. Approximately 8 hours after valve EMHV8823 had been declared inoperable, Callaway licensing personnel contacted the control room and informed them of an approved Technical Specification Bases change that did not allow Technical Specification 3.6.3 Condition C to be applicable to containment Penetration 49. The Technical Specification Bases change was effective May 1, 2008 but had not been issued to the control room. The licensee determined that the more restrictive Technical Specification 3.6.3, Condition A, should have been entered with an action to isolate the affected penetration within 4 hours. The licensee performed a containment entry following discovery of entry into Technical Specification 3.6.3, Condition A and found that Valve EMHV8823 failed its surveillance due to out of adjustment position indicator limit switches. The valve was verified closed and isolated allowing exit from Technical Specification 3.6.3, Condition A.
This finding, failure to ensure the Technical Specification Bases were maintained current and available to the Callaway control room staff, is more than minor because if left uncorrected, the failure to maintain the Technical Specification Bases current could become a more significant safety concern. This finding was determined to affect the barrier integrity cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding is determined to be of very low safety significance since this finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen ignitors in the reactor containment. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to communicate, in a timely manner, decisions to personnel who have a need to know the information in order to perform work safely [H.1(c)].
Inspection Report# : 2008003 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to comply wity high radiation area entry requirements.
The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.7.1, which resulted from a failure of three individuals to comply with high radiation area entry requirements. Specifically, on October 20, 2008, three engineers touring the reactor building entered a posted high radiation area without signing in on a radiation work permit which allowed entry into a high radiation area, and did not receive a briefing on dose rates in the high radiation area. Shortly after entering the high radiation area, one of the engineers received an electronic dosimeter rate alarm when dose rates in the area exceeded the 50 millirem per hour setpoint. The licensee entered this event into their
 
corrective action program and conducted an Event Review Team meeting to determine the probable causes that led to the event and recommend corrective actions to prevent the event from happening in the future.
Failure to comply with high radiation area entry requirements is a performance deficiency. This finding is greater than minor because it was associated with the cornerstone attribute of exposure control and affected the cornerstone objective, in that, the failure to meet high radiation area entry requirements increases the potential for increased radiation dose. This finding involved an individual workers' unplanned, unintended dose or potential of such dose (resulting from actions or conditions contrary to Technical Specifications) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised.
Additionally, the finding had a crosscutting aspect in the area of human performance, work practices component, because the workers failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2008005 (pdf)
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : April 07, 2009
 
Callaway 1Q/2009 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Response to Feedwater Transient Results in Reactor Trip The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after operator response to an electrical fault on the condensate Pump C motor resulted in an unplanned and unnecessary reactor trip, feedwater isolation, and auxiliary feedwater actuation. On December 11, 2008, Callaway Plant experienced an automatic turbine trip/reactor trip during a power reduction initiated by the operators response to a loss of condensate Pump C. The control room supervisor directed a power reduction without immediately referencing Procedure OTO AE 00001 guidance and without specifying any magnitude or rate limitations on the power reduction.
The balance of plant reactor operator, not aware of the procedural limitations, initiated the power reduction using the turbine controls load limiter potentiometer. This method of turbine load control eliminated all automatic rate-limiting functions. The steam generator levels increased rapidly with sluggish main feedwater regulating valves slowing anticipatory response. The steam generator P-14 high-high level turbine trip/reactor trip occurred about 5 minutes after condensate Pump C had tripped.
This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. The finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively establish clear expectations and standards regarding procedurally directed actions versus actions viewed as necessary to stabilize a plant transient.
Inspection Report# : 2009002 (pdf)
Significance:        Mar 24, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Intermediate Range Hi Flux Reactor Protection System Actuation W hile Shutdown The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after maintenance on intermediate range nuclear Instrument N36 resulted in an unanticipated reactor trip signal and feedwater isolation. On December 12, 2008, Callaway instrumentation and controls maintenance personnel performed work to replace a circuit card associated with the intermediate range nuclear Instrument P 6 bistable. At the time of the maintenance, the plant was in Mode 3 with the reactor trip breakers open. Shortly after beginning work, an intermediate range high flux reactor trip signal was generated. The trip signal was generated because the bypass of the reactor trip bistables is removed upon removal of the control power fuses. With instrument power removed, the solid state protection system perceived a high intermediate range neutron flux condition and generated a reactor trip signal and feedwater isolation. Control room operators responded to the feedwater isolation by starting both motor-driven auxiliary feedwater pumps and restoring steam generator water levels to the program band. The licensee later determined that instrumentation and controls maintenance personnel were unaware that pulling the control power fuses would cause a reactor trip signal and that the step in the work instruction that directed the removal of the control power fuses had not received an adequate review.
 
This finding was greater than minor because the finding impacted the Initiating Events cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This issue was entered into the licensee's corrective action program as Callaway Action Request 200812681. The finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to coordinate the impact of changes to the work scope or activity, specifically, the licensee failed to fully evaluate the impact of removal of control power fuses on the work instructions.
Inspection Report# : 2009002 (pdf)
Significance:        Feb 27, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Problems with Fire Protection Impairment Permits An NRC identified violation of License Condition 2.C.(5), Fire Protection, was identified for failing to effectively correct problems with the issuance and establishment of Fire Protection Impairment Permits. After problems were identified in 2006 and 2007, as a corrective action, the licensee conducted training in 2008 on the program requirements in the Maintenance and Operations Departments. Despite this corrective action, the licensee continued to experience failures to request a fire impairment and failures to implement pre-planned impairments. Some failures involved oversight problems for contract workers, who were not addressed in the training. Two procedural violations occurred in late 2008 that involved the failure to establish a Fire Protection Impairment Permit before performing hot work. The licensee has entered the issue into the corrective action program as Callaway Action Request (CAR) 200901638.
The inspectors determined that failing to correct problems associated with the use of required Fire Protection Impairment Permits is a performance deficiency. The finding is more than minor because it affects the protection against external factors attribute of the initiating events cornerstone, and it directly affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using the NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, Phase 1 worksheet, the finding was determined to be of very low safety significance (Green) because the condition represented a low degradation of fire prevention and administrative controls. The cause of the finding is related to the Human Performance cross-cutting component of Work Practices, in that the licensee failed to effectively communicate expectations and personnel failed to follow procedures [H.4.b].
Inspection Report# : 2009006 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to maintain an adequate plant shutdown procedure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after improper isolation of the main steam isolation valves by the Callaway control room operators resulted in a reactor trip signal and auxiliary feedwater actuation on October 11, 2008. Procedure OTG ZZ 00006, "Plant Cooldown Hot Standby to Cold Shutdown," allowed premature main steam isolation valve closures just after entering Mode 4. The operator then decided to reopen main steam isolation Valve A and atmospheric Steam Dump A. This created a significant increase in steam flow from the steam generator which caused the steam generator level to swell up to the P 14 steam generator high level feedwater isolation setpoint. The steam generator levels all decreased to the steam generator narrow range low-low setpoint generating the need for auxiliary feedwater actuation.
This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and
 
challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings," this finding is determined to be of very low safety significance since this finding did not affect the Technical Specification limit for reactor coolant system leakage, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to communicate, in a timely manner, decisions to personnel who have a need to know the information in order to perform work safely.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Failure to evaluate material equivalencies leads to a manual reactor trip The inspectors identified a self-revealing finding for failure of the engineering department to perform a material equivalency evaluation to ensure replacement components do not adversely affect plant operations. On November 11, 2008, Callaway Plant experienced a trip of main feedwater Pump B due to low lube oil pressure. Since the reactor was at greater than 80 percent power, the plant operators inserted a manual reactor trip. Following the reactor trip, maintenance personnel discovered two pieces of o-ring foreign material within main feedwater Pump B bearing oil supply pressure regulating Valve FCV0970. The foreign material was found wrapped around the regulating spring which inhibited valve movement and caused the lube oil low pressure condition. The licensee determined that the ethylene propylene diene M-class type o-ring became pliable when exposed to lube oil and was allowed to fall and be introduced into the system as foreign material. The ethylene propylene diene M-class o-rings had been approved as an equivalent replacement in July 1999 for the vendor recommended Buna-N type o-rings without performing an engineering material equivalency evaluation. Buna-N material is approved for use in petroleum based systems while ethylene propylene diene M-class is not.
This finding is greater than minor because it is associated with the design control attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using Manual Chapter 0609.04, "Phase 1 -
Initial Screening and Characterization of Findings," the finding is determined to be potentially risk significant because it contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. When evaluated per Manual Chapter 0609 Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," and the Callaway Plant Phase 2 pre-solved table item Failure to Reestablish Main Feedwater, the inspectors determined this finding to be of very low safety significance. This issue was entered into the licensee's corrective action program as Callaway Action Request 200811781. This finding was determined to not have a crosscutting aspect because the performance deficiency is not indicative of current licensee performance.
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate shutdown risk assessment for maintenance activities in the reactor building.
The inspectors identified a noncited violation of 10 CFR 50.65(a)(4), for failure to adequately assess and manage shutdown risk associated with maintenance activities in the reactor building. Specifically, on October 15, 2008, the inspectors found foreign material exclusion covers installed on the Train B containment recirculation sump. The covers were installed on October 14, 2008, per the direction of the containment coordinator without notification to the control room. The covers were installed to prevent debris from entering the sump. Following discussions with operations personnel, the inspectors found that the Train B containment recirculation sump was inappropriately
 
credited in the licensees shutdown safety assessment. An updated shutdown safety assessment was performed and it was determined that plant risk remained yellow.
This finding is greater than minor because the licensees risk assessment inappropriately credited risk-significant structures, systems and components that were unavailable during maintenance. This finding affected the Mitigating Systems cornerstone. Using Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the finding was found to be of very low safety significance because the licensee maintained two trains of decay heat removal operable and adequate equipment was available to support feed and bleed operations for at least 24 hours. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810540.
This finding had a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to obtain interdisciplinary input on safety-significant or risk-significant decisions. Specifically, the containment coordinator made a decision affecting the availability of the containment recirculation sumps without consulting the control room to determine the impact on plant risk.
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to ensure the suitability of the design of the resideual heat removal Train A pump room cooler The inspectors identified a self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after a trip of the residual heat removal Train A room cooler fan revealed that AmerenUE had not adequately selected and reviewed the suitability of the newly installed fan motor thermal overloads. Additionally, the NRC inspectors identified that the postmaintenance testing prescribed for the modified fan motor breaker did not allow sufficient time to challenge the thermal overload settings. On October 8, 2008, residual heat removal Train A room cooler fan shut down after only 22 minutes of run time. The breaker replacement modification used a calculation originally performed for the initial design of the old breaker which did not account for the cooler fan motor being a 20 horsepower motor nameplated down to a 10 horsepower rating.
This finding is greater than minor because it is similar to Manual Chapter 0612 "Examples of Minor Issues," Example 3j, in that the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of the component and a significant programmatic deficiency associated with postmaintenance test requirements was identified that could lead to worse errors if uncorrected. The inspectors determined that the finding impacted the Mitigating Systems cornerstone. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than Technical Specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810223. The inspectors determined that this finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action component because the AmerenUE modification for certain motor control center breakers failed to have a low enough threshold to identify fan motor rating and thermal overload setting errors.
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to adequately implement plant equipment control tagout procedure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after improper restoration of the essential service water supply to the emergency diesel generator Train A lubricating oil cooler resulted in significant water flow into the emergency diesel room on October 22, 2008. Two restoration evolutions associated with the essential service water and the emergency diesel generator systems had been proceeding in parallel. The reactor operator restoring the emergency diesel generator assumed the essential service
 
water supply was to remain isolated to the emergency diesel generator and thus changed the already approved worker protection assurance Clearance 71899 to leave the oil cooler drain valve open with no tag. Starting the essential service water pump pressurized the drain valve and produced significant water spray flow into the emergency diesel generator room until noticed by a diesel vendor representative about 30 minutes later.
This finding was greater than minor because if left uncorrected the deficiencies could become a more significant safety concern. The finding affected the Mitigating Systems cornerstone. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings," this finding is determined to be of very low safety significance since this finding was not a design or qualification deficiency, did not represent a loss of system or train safety function and did not screen as potentially risk significant due to a flooding initiating event using the criteria on the characterization worksheet. This finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee's pre-job briefing, self- and peer-checking, and proper documentation of activity were inadequate to overcome worker protection assurance clearance process problems and an inexperienced operating supervisor. These less than adequate worker practices resulted in personnel proceeding in the face of uncertainty.
Inspection Report# : 2008005 (pdf)
Significance: SL-IV Sep 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit a Licensee Event Report for a Condition Prohibitied by the Plant's Technical Specifications The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(1) for a failure to submit a required licensee event report within 60 days after discovery of an event requiring a report. On May 21, 2008, Callaway Plant personnel discovered a 6.6 cubic foot void of air within the safety injection system common suction piping. The voided piping, determined to have existed for over a year, was caused by relief valve maintenance on Valve EM8858A that occurred on May 7, 2007. Callaway Plant licensing staff performed a reportability evaluation and determined that the discovery of the void was not required to be reported to the NRC. The inspectors reviewed the licensees reportability evaluation and associated past operability and determined the event was reportable since a postulated single failure had the potential to disable both emergency core cooling system trains during cold leg recirculation. Since the emergency core cooling system was inoperable from May 7, 2007, until May 21, 2008, the event resulted in an operation or condition which was prohibited by the plants Technical Specifications as well as an event where a single cause or condition caused two independent trains to become inoperable in a single system.
This finding is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. This finding affected the mitigating systems cornerstone. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV, noncited violation. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810199. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate a void discovered in the emergency core cooling system for operability and reportability.
Inspection Report# : 2008004 (pdf)
Significance:      Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Safety Related 125 Vdc Station Battery NK11 Inadequate Battery Sizing Calculation The team identified a non-cited violation 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to verify the adequacy of design and for failure to correctly translate the 125 Vdc system design basis into instructions, procedures, and drawings. Specifically, the licensee failed to include momentary loads in the battery sizing calculation, thus reducing the peak load demand voltage during the first minute of an event, an intermediate scenario event, and the last minute of the battery duty cycle. Additionally, the licensees subsequent review determined that the
 
calculation had failed to include three additional momentary loads. The failure to include these loads prevented the licensee from developing a battery duty cycle profile that conforms to the guidance of IEEE 485-1983 and correctly simulates the battery loads following a design basis or station blackout event. The licensee entered this finding into their corrective action program as Callaway Action Request 200808609.
The failure to account for all loads, including momentary loads, in the battery design calculation was a performance deficiency because it prevented the licensee from correctly analyzing available voltage at safety-related components during the battery peak loading periods. The finding was more than minor because it is associated with the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the safety-related battery systems to respond to initiating events and prevent undesirable consequences.
Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because adequate margins had been included in the battery selection and, therefore, the issue was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2008008 (pdf)
Significance:      Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Pipe Break Location for the Condensate Stoarge Tank Supply to Auxiliary Feedwater Pumps The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for not verifying the adequacy of the design. Specifically, an incorrect pipe break location in the analysis for loss of the condensate storage tank feed to the auxiliary feedwater pumps caused the analysis to be non-conservative for the amount of water available to the auxiliary feedwater pumps. This error provided for more water to be available for use by the auxiliary feedwater pumps than would actually be available if the analysis had identified the correct location of the postulated pipe break. The licensee has entered this finding into their corrective action program as Callaway Action Request CAR 200808674.
The failure to meet design control requirements associated with the pipe break analysis with sufficient water to run the auxiliary feedwater pumps prior to switch over to the essential service water system is a performance deficiency. Per Manual Chapter 0612, Appendix E, Section 3, Non-significant Dimensional, Time, Calculation, or Drawing Discrepancies, Example J, this finding is more than minor because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system or component. Using Manual Chapter 0609, Significance Determination Process Phase 1 screening worksheet, the team determined that the finding was of very low safety significance. There was no actual loss of safety function and the new analysis demonstrated that the auxiliary feedwater pumps would have enough water available from the Condensate Storage Tank prior to switchover to the Essential Service Water system to complete their design function.
Inspection Report# : 2008008 (pdf)
Significance:      Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater Turbine Digital Control Panel FC219 The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawing for the failure to follow Procedure APA-ZZ-00500, Appendix 1, Revision 6, Operability Determination.
The evaluation did not include the additional heat loading on equipment in the turbine driven auxiliary feedwater pump room, caused from an active steam leak from the turbine governor end case joint. The licensee had failed to include the additional steam leak heat load in either of the room temperature calculations M-GF-415 or BO -05, which were used in the operability determination. The heat input into the room, due to the steam leak, may have adversely affected the operation of the turbine digital speed control unit. The licensee has entered this finding into their corrective action program as Callaway Action Request 200808777.
 
The failure to either correct the active steam leak or to account for the leak in their design calculations, is a performance deficiency. Per Inspection Manual Chapter 0612, Appendix E, Section 3, Non-significant Dimensional, Time, Calculation, or Drawing Discrepancies, Example J, this finding is more than minor because the licensee had not resolved the deficiency, resulting in a condition in which there was a reasonable doubt regarding the reliability of the turbine digital speed control unit. Using Inspection Manual Chapter 0609, "Significance Determination Process,"
Phase 1 screening worksheets, the team determined that the finding was of very low safety significance. Since there was no actual loss of safety function and the new analysis demonstrated that the maximum room temperature, including the additional heat load, would not exceed the design limit of digital turbine speed controls unit, the issue was a design deficiency confirmed not to result in loss of operability per NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The finding had crosscutting aspects in the area of human performance (decision making) because the licensee used non-conservative assumptions in decision making and failed to either repair the active steam leak, or to account for it in their design calculations. This activity was indicative of current performance as the steam leak still existed and had not been included in the design calculations until October 2008.
Inspection Report# : 2008008 (pdf)
Significance:        Jun 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Surveillance Procedure Resulted in an Inaperable Emergency Core Cooling System The inspectors identified a noncited violation of Technical Specification 3.5.2, "Emergency Core Cooling Systems,"
after an inadequate surveillance procedure resulted in the licensee failing to maintain the emergency core cooling system full of water as required per Technical Specification 3.5.2. On May 21, 2008, Callaway Plant engineering discovered that a section of the cold leg recirculation piping, specifically the discharge of the residual heat removal pumps to the safety injection pumps, contained 6.6 cubic feet of air. Callaway monthly surveillance Procedure OSP SA 00003, "Emergency Core Cooling Flow Path Verification and Venting," had a purpose to: "Verify the ECCS is full of water," in accordance with Technical Specification Surveillance Requirement 3.5.2.3. The monthly verification and vent procedure was not comprehensive enough to ensure all the emergency core cooling system was full of water.
This finding is more than minor because it was similar to Example 3e of NRC Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," and met the Not Minor If, criteria because the failure to meet the licensees administrative requirement for allowable void fraction impacted the ability of the Train A safety injection system to function upon initiation of high-pressure recirculation. This finding affected the mitigating systems cornerstone procedure quality attribute. Using the Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding should be evaluated using the Phase 2 process described in Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. As described in Section III, of Appendix A, given that the presolved table did not contain a suitable target or surrogate for this finding, the senior reactor analyst used the risk-informed notebook to evaluate the significance of this finding affecting only high-pressure recirculation as very low risk significance (Green). This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions in decision making and did not adopt a requirement to demonstrate that a single vent valve was sufficient to vent the affected line rather than assuming that an additional installed valve was not necessary to completely fill, vent, and test the line [H.1(b)].
Inspection Report# : 2008003 (pdf)
Significance:        Jun 24, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct a Condition Adverse to Quality for Diesel Generator Jacket Water O-Rings A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified after the licensee failed to promptly correct leakage from diesel generator jacket water o-rings. On February 20, 2008, during a normal surveillance run of Emergency Diesel Generator B, Callaway operations personnel identified an approximately 80 drops per minute jacket water leak caused by premature failure of Nitrile type o-rings.
 
Following restoration of Emergency Diesel Generator B, the licensee re-evaluated the preventative maintenance frequency for jacket water o-ring replacement and reduced the replacement frequency from once every three years to once every refueling cycle. Then, on May 28, 2008, during a routine surveillance run of Emergency Diesel Generator A, Callaway operations personnel identified that Emergency Diesel Generator A had a 200 drops per minute jacket water leak. Similar to the condition observed on Emergency Diesel Generator B on February 20, 2008, the source of the leakage was from Nitrile type o-rings within the jacket water system. The o-rings responsible for jacket water leakage were found to be of similar age to those that failed during the February 20, 2008 surveillance but had not been replaced despite the change to the licensee's preventive maintenance frequency.
This finding, failure to implement adequate corrective actions for degraded Nitrile type o-rings in Emergency Diesel Generator A after previously identifying the adverse condition on Emergency Diesel Generator B, is more than minor because, if left uncorrected, degraded diesel generator jacket water o-rings could become a more significant safety concern. This finding affected the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in loss of operability. This finding has a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to plan work activities to support long-term equipment reliability by addressing known degraded conditions in a more reactive than preventative manner [H.3(b)].
Inspection Report# : 2008003 (pdf)
Significance:      Jun 24, 2008 Identified By: NRC Item Type: VIO Violation Failure to Prevent Recurrence of Voids in Emergency Core Cooling System Cold Leg Recirculation Piping The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," because the licensee failed to restore compliance within a reasonable time by establishing measures to prevent void formation in emergency core cooling system suction piping for the Train A safety injection system. On May 21, 2008, Callaway Plant engineering performed ultrasonic inspection of the safety injection system common suction piping Line EM¬023 HCB - 6" and discovered a 6.6 cubic foot voided area. This exceeded the allowable void fraction of 2.1 cubic feet required for operability. This voided piping, determined to have existed for over a year, was caused by relief valve maintenance on Valve EM8858A (May 7, 2007). The maintenance restoration failed to perform a fill and vent to ensure the suction pipe was full of water. The inspectors identified several related examples where the licensee had performed either inadequate operating experience evaluations, inadequate extent of condition reviews, or inadequate procedure corrections.
This finding, failure to restore compliance to prevent recurrence of emergency core cooling system voids was more than minor because it is similar to Example 3e of NRC Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," criteria because the failure impacted the ability of the emergency core cooling system to function upon initiation of high-pressure recirculation. Using the Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding should be evaluated using the Phase 2 process described in Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. As described in Section III, of Appendix A, given that the presolved table did not contain a suitable target or surrogate for this finding, the senior reactor analyst used the risk-informed notebook to evaluate the significance of this finding as very low risk significance (Green). This finding has a crosscutting aspect in the area of problem identification and resolution associated witht the corrective action program component because AmerenUE failed to thoroughly evaluate voiding problems such that the resolutions addressed causes and extent of condition, as necessary [P.1(c)].
Inspection Report# : 2008005 (pdf)
Inspection Report# : 2008003 (pdf)
Barrier Integrity
 
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate maintenance procedure resulted in residual heat removal mechanical seal failure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1a, Procedures, for inadequate procedural guidance that resulted in the failure of the residual heat removal Train A pump mechanical seal.
On October 22, 2008, the licensee discovered a solid stream of water issuing from the residual heat removal Train A pump mechanical seal. The failure occurred because of installation difficulties encountered on October 8, 2008, when the seal sleeve was installed with the seal locking collar engaged. This configuration resulted in increased loading on the seal seating surfaces that resulted in surface chipping and led to seal failure after approximately 48 hours of shutdown cooling operation. Mechanical seal replacement Procedure MPM EJ QP001, Residual Heat Removal Pump Overhaul, did not specify that the seal sleeve needed to be installed prior to installing the seal-locking collar.
Additionally, the installation procedure did not specify any post-installation acceptance criteria to ensure the seal is properly seated. An analysis of the seal failure determined that leakage would not exceed the 2 gallon per minute Technical Specification limit but would exceed the 1 gallon per minute administrative limit for emergency core cooling system leakage outside containment.
This finding is more than minor because it was associated with the Barrier Integrity cornerstone attribute of procedural quality and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process," this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since the 2 gallon per minute limit assumed in the post accident dose calculation was preserved and therefore the degraded condition would have no impact on large early release frequency. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810933. This finding did not have a crosscutting aspect since it was not a performance deficiency indicative of current licensee performance.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to terminate refueling water storage tank recirculation results in inadvertent loss of spent fuel pool inventory The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1a, Procedures, for the failure to close Valve BNV0002 during a fill of the spent fuel pool resulting in approximately 2000 gallons of water being inadvertently transferred from the spent fuel pool to the refueling water storage tank. On November 7, 2008, Procedure OTN EC 00001 was performed to add makeup water to the spent fuel pool. Prior to performing the evolution, operations briefed that the refueling water storage tank was on recirculation and that this alignment needed to be secured prior to performing a fill of the spent fuel pool. Following termination of the refueling water storage tank recirculation lineup and after a fill of the spent fuel pool was initiated, the control room received annunciator RWST Lev HILO. The crew recognized that an inadvertent transfer of spent fuel pool water to the refueling water storage tank was in progress and directed that Valves ECV0076 and BNV0002 be closed. It was later discovered that poor communication between operators on the status of Valve BNV0002 resulted in the refueling water storage tank remaining on recirculation during the fill operation.
This finding is more than minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because it only represents a degradation of the radiological barrier function provided by the spent fuel pool. This issue was entered into the licensee's corrective action program as Callaway Action Request 200811692.
This finding had a crosscutting aspect in the area of human performance associated with the work control component
 
because operations personnel failed to effectively communicate work status to the control room.
Inspection Report# : 2008005 (pdf)
Significance:      Sep 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Boric Acid Corrosion Control Procedures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform a corrosion evaluation of boric acid leakage from containment spray Valve ENHV0006 in accordance with Procedure EDP ZZ 01004, Boric Acid Corrosion Control Program. On August 29, 2008, the resident inspectors identified an active packing leak on Valve ENHV0006 with impact to carbon steel components on the valve as evident by discolored, brown boron. The leak, which had been active since February 27, 2007, was caused by a stem imperfection that was previously identified on December 5, 2007. The inspectors noted that Valve ENHV0006 did not have a current boric acid corrosion evaluation despite meeting the screening requirements for an evaluation listed in Procedure EDP ZZ 01004, Boric Acid Corrosion Control Program, Section 4.2. Programmatic boric acid control and work control issues were a key contributor to not recognizing the need for an updated boric acid corrosion evaluation.
This finding is more than minor because, if left uncorrected, the failure to analyze the effects of boric acid corrosion on safety related components could become a more significant safety concern. This finding affected the barrier integrity cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined be of very low safety significance because the finding does not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere, does not represent an actual open pathway in the physical integrity of the reactor containment, and does not involve an actual reduction in function of hydrogen ignitors in the reactor containment. This issue was entered into the licensee's corrective action program as Callaway Action Request 200809351. This finding has a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to interdepartmentally coordinate the impact of changes to the work scope for Valve ENHV0006 such that appropriate personnel could perform the necessary evaluations to assure plant performance.
Inspection Report# : 2008004 (pdf)
Significance:      Jun 24, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Ensure the Suitability of the Design of the Containment Air Cooler Control Circuitry A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified after determining that the licensee had not adequately selected and reviewed the suitability of the design of the containment air cooler control circuitry. On March 26, 2008, containment air Cooler A fan shut down when shifted from fast to slow speed. Troubleshooting by the licensee determined that voltage was lost to the control power circuitry when the fast speed thermal overload tripped. Since the overload contacts were wired in series, containment air Cooler A experienced a complete loss of control power rendering it inoperable. The licensee determined the trip to be caused by operation of containment air coolers in fast speed, during a period of higher than normal containment pressure. The licensee analyzed the potential impact of the newly discovered adverse containment cooler design vulnerability against design basis accident scenarios. The licensee determined that a hot zero power main steam line break results in a delayed safety injection signal allowing the fan motor overloads to trip prior to being shed by the load sequencer. The containment air coolers would then experience a complete loss of control power and would not be capable of automatically restarting in slow speed. The analysis revealed that the peak containment pressure limit of 48.1 psig would be preserved. The licensee submitted a Licensee Event Report as required by 10 CFR 50.73 since the inadequate containment air cooler control circuitry resulted in a condition prohibited by the plants Technical Specifications.
This finding, failure to ensure the design of the containment air cooler control circuitry was suitable for all plant conditions, was more than minor because it was associated with the barrier integrity cornerstone attribute of design
 
control and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609 Appendix H, Containment Integrity Significance Determination Process," this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since containment coolers are structures, systems or components that have no impact on large early release frequency. The inspectors determined that this finding does not have a crosscutting aspect associated with it since the performance deficiency was not indicative of current licensee performance.
Inspection Report# : 2008003 (pdf)
Significance:      Jun 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Technical Specificaion Bases Change Process The inspectors identified a noncited violation of Technical Specification 5.4.1.a, Procedures, after Callaway control room operators improperly entered a wrong Technical Specification action statement due to the failure to maintain the Technical Specification Bases current. On June 17, 2008, during surveillance testing, Valve EMHV8823 failed to indicate fully closed. Since EMHV8823 is an isolation valve for containment Penetration 49, the licensee entered Technical Specification 3.6.3, Containment Isolation Valves, Condition C, with an action to restore the valve to an operable status or isolate the penetration within 72 hours. Approximately 8 hours after valve EMHV8823 had been declared inoperable, Callaway licensing personnel contacted the control room and informed them of an approved Technical Specification Bases change that did not allow Technical Specification 3.6.3 Condition C to be applicable to containment Penetration 49. The Technical Specification Bases change was effective May 1, 2008 but had not been issued to the control room. The licensee determined that the more restrictive Technical Specification 3.6.3, Condition A, should have been entered with an action to isolate the affected penetration within 4 hours. The licensee performed a containment entry following discovery of entry into Technical Specification 3.6.3, Condition A and found that Valve EMHV8823 failed its surveillance due to out of adjustment position indicator limit switches. The valve was verified closed and isolated allowing exit from Technical Specification 3.6.3, Condition A.
This finding, failure to ensure the Technical Specification Bases were maintained current and available to the Callaway control room staff, is more than minor because if left uncorrected, the failure to maintain the Technical Specification Bases current could become a more significant safety concern. This finding was determined to affect the barrier integrity cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding is determined to be of very low safety significance since this finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen ignitors in the reactor containment. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to communicate, in a timely manner, decisions to personnel who have a need to know the information in order to perform work safely [H.1(c)].
Inspection Report# : 2008003 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to comply wity high radiation area entry requirements.
 
The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.7.1, which resulted from a failure of three individuals to comply with high radiation area entry requirements. Specifically, on October 20, 2008, three engineers touring the reactor building entered a posted high radiation area without signing in on a radiation work permit which allowed entry into a high radiation area, and did not receive a briefing on dose rates in the high radiation area. Shortly after entering the high radiation area, one of the engineers received an electronic dosimeter rate alarm when dose rates in the area exceeded the 50 millirem per hour setpoint. The licensee entered this event into their corrective action program and conducted an Event Review Team meeting to determine the probable causes that led to the event and recommend corrective actions to prevent the event from happening in the future.
Failure to comply with high radiation area entry requirements is a performance deficiency. This finding is greater than minor because it was associated with the cornerstone attribute of exposure control and affected the cornerstone objective, in that, the failure to meet high radiation area entry requirements increases the potential for increased radiation dose. This finding involved an individual workers' unplanned, unintended dose or potential of such dose (resulting from actions or conditions contrary to Technical Specifications) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised.
Additionally, the finding had a crosscutting aspect in the area of human performance, work practices component, because the workers failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2008005 (pdf)
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : June 05, 2009
 
Callaway 2Q/2009 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Response to Feedwater Transient Results in Reactor Trip The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after operator response to an electrical fault on the condensate Pump C motor resulted in an unplanned and unnecessary reactor trip, feedwater isolation, and auxiliary feedwater actuation. On December 11, 2008, Callaway Plant experienced an automatic turbine trip/reactor trip during a power reduction initiated by the operators response to a loss of condensate Pump C. The control room supervisor directed a power reduction without immediately referencing Procedure OTO AE 00001 guidance and without specifying any magnitude or rate limitations on the power reduction.
The balance of plant reactor operator, not aware of the procedural limitations, initiated the power reduction using the turbine controls load limiter potentiometer. This method of turbine load control eliminated all automatic rate-limiting functions. The steam generator levels increased rapidly with sluggish main feedwater regulating valves slowing anticipatory response. The steam generator P-14 high-high level turbine trip/reactor trip occurred about 5 minutes after condensate Pump C had tripped.
This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. The finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively establish clear expectations and standards regarding procedurally directed actions versus actions viewed as necessary to stabilize a plant transient.
Inspection Report# : 2009002 (pdf)
Significance:        Mar 24, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Intermediate Range Hi Flux Reactor Protection System Actuation W hile Shutdown The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after maintenance on intermediate range nuclear Instrument N36 resulted in an unanticipated reactor trip signal and feedwater isolation. On December 12, 2008, Callaway instrumentation and controls maintenance personnel performed work to replace a circuit card associated with the intermediate range nuclear Instrument P 6 bistable. At the time of the maintenance, the plant was in Mode 3 with the reactor trip breakers open. Shortly after beginning work, an intermediate range high flux reactor trip signal was generated. The trip signal was generated because the bypass of the reactor trip bistables is removed upon removal of the control power fuses. With instrument power removed, the solid state protection system perceived a high intermediate range neutron flux condition and generated a reactor trip signal and feedwater isolation. Control room operators responded to the feedwater isolation by starting both motor-driven auxiliary feedwater pumps and restoring steam generator water levels to the program band. The licensee later determined that instrumentation and controls maintenance personnel were unaware that pulling the control power fuses would cause a reactor trip signal and that the step in the work instruction that directed the removal of the control power fuses had not received an adequate review.
 
This finding was greater than minor because the finding impacted the Initiating Events cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This issue was entered into the licensee's corrective action program as Callaway Action Request 200812681. The finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to coordinate the impact of changes to the work scope or activity, specifically, the licensee failed to fully evaluate the impact of removal of control power fuses on the work instructions.
Inspection Report# : 2009002 (pdf)
Significance:        Feb 27, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Problems with Fire Protection Impairment Permits An NRC identified violation of License Condition 2.C.(5), Fire Protection, was identified for failing to effectively correct problems with the issuance and establishment of Fire Protection Impairment Permits. After problems were identified in 2006 and 2007, as a corrective action, the licensee conducted training in 2008 on the program requirements in the Maintenance and Operations Departments. Despite this corrective action, the licensee continued to experience failures to request a fire impairment and failures to implement pre-planned impairments. Some failures involved oversight problems for contract workers, who were not addressed in the training. Two procedural violations occurred in late 2008 that involved the failure to establish a Fire Protection Impairment Permit before performing hot work. The licensee has entered the issue into the corrective action program as Callaway Action Request (CAR) 200901638.
The inspectors determined that failing to correct problems associated with the use of required Fire Protection Impairment Permits is a performance deficiency. The finding is more than minor because it affects the protection against external factors attribute of the initiating events cornerstone, and it directly affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using the NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, Phase 1 worksheet, the finding was determined to be of very low safety significance (Green) because the condition represented a low degradation of fire prevention and administrative controls. The cause of the finding is related to the Human Performance cross-cutting component of Work Practices, in that the licensee failed to effectively communicate expectations and personnel failed to follow procedures [H.4.b].
Inspection Report# : 2009006 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to maintain an adequate plant shutdown procedure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after improper isolation of the main steam isolation valves by the Callaway control room operators resulted in a reactor trip signal and auxiliary feedwater actuation on October 11, 2008. Procedure OTG ZZ 00006, "Plant Cooldown Hot Standby to Cold Shutdown," allowed premature main steam isolation valve closures just after entering Mode 4. The operator then decided to reopen main steam isolation Valve A and atmospheric Steam Dump A. This created a significant increase in steam flow from the steam generator which caused the steam generator level to swell up to the P 14 steam generator high level feedwater isolation setpoint. The steam generator levels all decreased to the steam generator narrow range low-low setpoint generating the need for auxiliary feedwater actuation.
This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and
 
challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings," this finding is determined to be of very low safety significance since this finding did not affect the Technical Specification limit for reactor coolant system leakage, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to communicate, in a timely manner, decisions to personnel who have a need to know the information in order to perform work safely.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Failure to evaluate material equivalencies leads to a manual reactor trip The inspectors identified a self-revealing finding for failure of the engineering department to perform a material equivalency evaluation to ensure replacement components do not adversely affect plant operations. On November 11, 2008, Callaway Plant experienced a trip of main feedwater Pump B due to low lube oil pressure. Since the reactor was at greater than 80 percent power, the plant operators inserted a manual reactor trip. Following the reactor trip, maintenance personnel discovered two pieces of o-ring foreign material within main feedwater Pump B bearing oil supply pressure regulating Valve FCV0970. The foreign material was found wrapped around the regulating spring which inhibited valve movement and caused the lube oil low pressure condition. The licensee determined that the ethylene propylene diene M-class type o-ring became pliable when exposed to lube oil and was allowed to fall and be introduced into the system as foreign material. The ethylene propylene diene M-class o-rings had been approved as an equivalent replacement in July 1999 for the vendor recommended Buna-N type o-rings without performing an engineering material equivalency evaluation. Buna-N material is approved for use in petroleum based systems while ethylene propylene diene M-class is not.
This finding is greater than minor because it is associated with the design control attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using Manual Chapter 0609.04, "Phase 1 -
Initial Screening and Characterization of Findings," the finding is determined to be potentially risk significant because it contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. When evaluated per Manual Chapter 0609 Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," and the Callaway Plant Phase 2 pre-solved table item Failure to Reestablish Main Feedwater, the inspectors determined this finding to be of very low safety significance. This issue was entered into the licensee's corrective action program as Callaway Action Request 200811781. This finding was determined to not have a crosscutting aspect because the performance deficiency is not indicative of current licensee performance.
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance: SL-IV Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit Complete and Accurate Risk Information for a Requested License Amendment The inspectors identified a noncited violation of 10 CFR 50.9, "Completeness and Accuracy of Information," when AmerenUE failed to submit complete and accurate quantification of risk contributors associated with a license amendment supporting a modification to replace the underground portion of the essential service water system Train B piping with high density polyethelene pipe. The inspectors questioned the risk impact of a possible control room fire which led to the discovery that the licensee had not followed their process for screening out fire areas. The licensee entered this item into their corrective action program as Callaway Action Request 200902810 and also submitted an update to License Amendment 191 to correctly account for the control room fire risk.
 
This finding affects the Mitigating Systems cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect because the licensees failure to thoroughly review and submit the risk for control room fires was not part of a corrective action process, but instead an oversight by the licensing review process.
Inspection Report# : 2009003 (pdf)
Significance:      Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Controls of Crane Work Above the Protected Train of Essential Service Water The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) associated with the licensees failure to adequately assess and manage risk associated with crane work over the essential service water system Train A. On March 31, 2009, the licensee performed work in the vicinity of the protected essential service water system train which included movement of 1800 pound sand bags over the protected train piping. After questioning by the resident inspectors, the licensee determined that the lifts were not conducted in accordance with station procedures since the requirements of a required engineering judgment memo were not translated into work documents. The licensee entered this item into their corrective action program as Callaway Action Request 200902726.
The finding affected the Mitigating Systems cornerstone and was determined to be more than minor because the licensee failed to implement the prescribed significant compensatory measures associated with crane work in the vicinity of safe shutdown equipment. This finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to include appropriate risk insights in planned work activities.
Inspection Report# : 2009003 (pdf)
Significance:      Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate, At Power, Risk Assessment for Maintenance Activities on One Train of Essential Service Water and Emergency Diesel Generator The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) associated with the licensees failure to perform an adequate risk assessment for planned maintenance on the emergency diesel generator Train A and essential service water pump Train A. On April 28, 2009, Callaway Plant operators removed the emergency diesel generator Train A and essential service water pump Train A from service. The inspectors' review of the plant risk profile for the in-progress maintenance activity uncovered that this risk had not been accounted for by the plant safety monitor tool.
The licensee entered this item into their corrective action program as Callaway Action Request 200903480 The finding is more than minor because the risk, when correctly assessed, put the plant into a higher risk category for large early release frequency. Also the licensee risk assessment failed to consider risk significant systems, structures, and components and support systems that were unavailable during the maintenance. This finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to appropriately plan work activities consistent with nuclear safety by incorporating risk insights.
Inspection Report# : 2009003 (pdf)
Significance:      Apr 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Suitable Replacement Parts Essential for Emergency Diesel Generator Train B
 
The inspectors identifed a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" after the licensee failed to adequately select suitable replacement parts essential to the operation of emergency diesel generator Train B. On December 24, 2008, during performance of Procedure OSP-NE-0001B, "Standby Diesel Generator B Periodic Tests, " Callaway operations personnel identified that the emergency diesel generator Train B had an approximately 0.82 gallon per minute jacket water leak resulting in operators declaring the equipment inoperable.
Upon removal, the gasket was found to be soft and extruding from the flange edge. The licensee originally concluded the gasket failed due to vibrations associated with engine shutdown but altered that conclusion after discussions witht he resident inspectors and additional investigation. The licensee ultimately determined that the cause of the failure was due to incorrect gasket material being used during Job W200773 performed on October 16, 1999. The gasket was 1/8" thick which resulted ina lack of compression. Since the gaskets are composed of an aramid fiberous material, the lack of compression allowed the gasket to absorb water and soften. The leak identified on December 24, 2008, developed once the gasket softened sufficiently to extrude from the flange edge. This issue has been entered into the licensee's corrective action program as Callaway Action Request 200812985.
This finding was greater than minor because it was associated with the mitigating systems cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings, " this finding was determined to represent an actual loss of safety function of a single train for greater thatn its Technical Specification allowed outage time. When evaluated per Manual Chapter 0609 Appendiz A, "Determining the Significance of Reactor Inspection Finding for At-Power Situations," and the Callaway Plant P hae 2 pre-solved table item "Diesel Generator Fails to Run after Start," the inspectors determined this finding to be potentially risk significant. This finding was forwarded to a senior reactor analyst for review. The results of the senior reactor analyst's Phase 3 analysis determined the finding to be of very low safety significance. This finding did not have a crosscutting aspect since it was not a performance deficiency indicative of current licensee performance.
Inspection Report# : 2009007 (pdf)
Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate shutdown risk assessment for maintenance activities in the reactor building.
The inspectors identified a noncited violation of 10 CFR 50.65(a)(4), for failure to adequately assess and manage shutdown risk associated with maintenance activities in the reactor building. Specifically, on October 15, 2008, the inspectors found foreign material exclusion covers installed on the Train B containment recirculation sump. The covers were installed on October 14, 2008, per the direction of the containment coordinator without notification to the control room. The covers were installed to prevent debris from entering the sump. Following discussions with operations personnel, the inspectors found that the Train B containment recirculation sump was inappropriately credited in the licensees shutdown safety assessment. An updated shutdown safety assessment was performed and it was determined that plant risk remained yellow.
This finding is greater than minor because the licensees risk assessment inappropriately credited risk-significant structures, systems and components that were unavailable during maintenance. This finding affected the Mitigating Systems cornerstone. Using Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the finding was found to be of very low safety significance because the licensee maintained two trains of decay heat removal operable and adequate equipment was available to support feed and bleed operations for at least 24 hours. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810540.
This finding had a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to obtain interdisciplinary input on safety-significant or risk-significant decisions. Specifically, the containment coordinator made a decision affecting the availability of the containment recirculation sumps without consulting the control room to determine the impact on plant risk.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008
 
Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to ensure the suitability of the design of the resideual heat removal Train A pump room cooler The inspectors identified a self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after a trip of the residual heat removal Train A room cooler fan revealed that AmerenUE had not adequately selected and reviewed the suitability of the newly installed fan motor thermal overloads. Additionally, the NRC inspectors identified that the postmaintenance testing prescribed for the modified fan motor breaker did not allow sufficient time to challenge the thermal overload settings. On October 8, 2008, residual heat removal Train A room cooler fan shut down after only 22 minutes of run time. The breaker replacement modification used a calculation originally performed for the initial design of the old breaker which did not account for the cooler fan motor being a 20 horsepower motor nameplated down to a 10 horsepower rating.
This finding is greater than minor because it is similar to Manual Chapter 0612 "Examples of Minor Issues," Example 3j, in that the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of the component and a significant programmatic deficiency associated with postmaintenance test requirements was identified that could lead to worse errors if uncorrected. The inspectors determined that the finding impacted the Mitigating Systems cornerstone. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than Technical Specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810223. The inspectors determined that this finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action component because the AmerenUE modification for certain motor control center breakers failed to have a low enough threshold to identify fan motor rating and thermal overload setting errors.
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to adequately implement plant equipment control tagout procedure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after improper restoration of the essential service water supply to the emergency diesel generator Train A lubricating oil cooler resulted in significant water flow into the emergency diesel room on October 22, 2008. Two restoration evolutions associated with the essential service water and the emergency diesel generator systems had been proceeding in parallel. The reactor operator restoring the emergency diesel generator assumed the essential service water supply was to remain isolated to the emergency diesel generator and thus changed the already approved worker protection assurance Clearance 71899 to leave the oil cooler drain valve open with no tag. Starting the essential service water pump pressurized the drain valve and produced significant water spray flow into the emergency diesel generator room until noticed by a diesel vendor representative about 30 minutes later.
This finding was greater than minor because if left uncorrected the deficiencies could become a more significant safety concern. The finding affected the Mitigating Systems cornerstone. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings," this finding is determined to be of very low safety significance since this finding was not a design or qualification deficiency, did not represent a loss of system or train safety function and did not screen as potentially risk significant due to a flooding initiating event using the criteria on the characterization worksheet. This finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee's pre-job briefing, self- and peer-checking, and proper documentation of activity were inadequate to overcome worker protection assurance clearance process problems and an inexperienced operating supervisor. These less than adequate worker practices resulted in personnel proceeding in the face of uncertainty.
Inspection Report# : 2008005 (pdf)
 
Significance: SL-IV Sep 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit a Licensee Event Report for a Condition Prohibitied by the Plant's Technical Specifications The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(1) for a failure to submit a required licensee event report within 60 days after discovery of an event requiring a report. On May 21, 2008, Callaway Plant personnel discovered a 6.6 cubic foot void of air within the safety injection system common suction piping. The voided piping, determined to have existed for over a year, was caused by relief valve maintenance on Valve EM8858A that occurred on May 7, 2007. Callaway Plant licensing staff performed a reportability evaluation and determined that the discovery of the void was not required to be reported to the NRC. The inspectors reviewed the licensees reportability evaluation and associated past operability and determined the event was reportable since a postulated single failure had the potential to disable both emergency core cooling system trains during cold leg recirculation. Since the emergency core cooling system was inoperable from May 7, 2007, until May 21, 2008, the event resulted in an operation or condition which was prohibited by the plants Technical Specifications as well as an event where a single cause or condition caused two independent trains to become inoperable in a single system.
This finding is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. This finding affected the mitigating systems cornerstone. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV, noncited violation. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810199. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate a void discovered in the emergency core cooling system for operability and reportability.
Inspection Report# : 2008004 (pdf)
Significance:      Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Safety Related 125 Vdc Station Battery NK11 Inadequate Battery Sizing Calculation The team identified a non-cited violation 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to verify the adequacy of design and for failure to correctly translate the 125 Vdc system design basis into instructions, procedures, and drawings. Specifically, the licensee failed to include momentary loads in the battery sizing calculation, thus reducing the peak load demand voltage during the first minute of an event, an intermediate scenario event, and the last minute of the battery duty cycle. Additionally, the licensees subsequent review determined that the calculation had failed to include three additional momentary loads. The failure to include these loads prevented the licensee from developing a battery duty cycle profile that conforms to the guidance of IEEE 485-1983 and correctly simulates the battery loads following a design basis or station blackout event. The licensee entered this finding into their corrective action program as Callaway Action Request 200808609.
The failure to account for all loads, including momentary loads, in the battery design calculation was a performance deficiency because it prevented the licensee from correctly analyzing available voltage at safety-related components during the battery peak loading periods. The finding was more than minor because it is associated with the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the safety-related battery systems to respond to initiating events and prevent undesirable consequences.
Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because adequate margins had been included in the battery selection and, therefore, the issue was a design deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment.
Inspection Report# : 2008008 (pdf)
 
Significance:        Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Pipe Break Location for the Condensate Stoarge Tank Supply to Auxiliary Feedwater Pumps The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for not verifying the adequacy of the design. Specifically, an incorrect pipe break location in the analysis for loss of the condensate storage tank feed to the auxiliary feedwater pumps caused the analysis to be non-conservative for the amount of water available to the auxiliary feedwater pumps. This error provided for more water to be available for use by the auxiliary feedwater pumps than would actually be available if the analysis had identified the correct location of the postulated pipe break. The licensee has entered this finding into their corrective action program as Callaway Action Request CAR 200808674.
The failure to meet design control requirements associated with the pipe break analysis with sufficient water to run the auxiliary feedwater pumps prior to switch over to the essential service water system is a performance deficiency. Per Manual Chapter 0612, Appendix E, Section 3, Non-significant Dimensional, Time, Calculation, or Drawing Discrepancies, Example J, this finding is more than minor because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system or component. Using Manual Chapter 0609, Significance Determination Process Phase 1 screening worksheet, the team determined that the finding was of very low safety significance. There was no actual loss of safety function and the new analysis demonstrated that the auxiliary feedwater pumps would have enough water available from the Condensate Storage Tank prior to switchover to the Essential Service Water system to complete their design function.
Inspection Report# : 2008008 (pdf)
Significance:        Aug 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Auxiliary Feedwater Turbine Digital Control Panel FC219 The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawing for the failure to follow Procedure APA-ZZ-00500, Appendix 1, Revision 6, Operability Determination.
The evaluation did not include the additional heat loading on equipment in the turbine driven auxiliary feedwater pump room, caused from an active steam leak from the turbine governor end case joint. The licensee had failed to include the additional steam leak heat load in either of the room temperature calculations M-GF-415 or BO -05, which were used in the operability determination. The heat input into the room, due to the steam leak, may have adversely affected the operation of the turbine digital speed control unit. The licensee has entered this finding into their corrective action program as Callaway Action Request 200808777.
The failure to either correct the active steam leak or to account for the leak in their design calculations, is a performance deficiency. Per Inspection Manual Chapter 0612, Appendix E, Section 3, Non-significant Dimensional, Time, Calculation, or Drawing Discrepancies, Example J, this finding is more than minor because the licensee had not resolved the deficiency, resulting in a condition in which there was a reasonable doubt regarding the reliability of the turbine digital speed control unit. Using Inspection Manual Chapter 0609, "Significance Determination Process,"
Phase 1 screening worksheets, the team determined that the finding was of very low safety significance. Since there was no actual loss of safety function and the new analysis demonstrated that the maximum room temperature, including the additional heat load, would not exceed the design limit of digital turbine speed controls unit, the issue was a design deficiency confirmed not to result in loss of operability per NRC Manual Chapter Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The finding had crosscutting aspects in the area of human performance (decision making) because the licensee used non-conservative assumptions in decision making and failed to either repair the active steam leak, or to account for it in their design calculations. This activity was indicative of current performance as the steam leak still existed and had not been included in the design calculations until October 2008.
Inspection Report# : 2008008 (pdf)
 
Barrier Integrity Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate maintenance procedure resulted in residual heat removal mechanical seal failure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1a, Procedures, for inadequate procedural guidance that resulted in the failure of the residual heat removal Train A pump mechanical seal.
On October 22, 2008, the licensee discovered a solid stream of water issuing from the residual heat removal Train A pump mechanical seal. The failure occurred because of installation difficulties encountered on October 8, 2008, when the seal sleeve was installed with the seal locking collar engaged. This configuration resulted in increased loading on the seal seating surfaces that resulted in surface chipping and led to seal failure after approximately 48 hours of shutdown cooling operation. Mechanical seal replacement Procedure MPM EJ QP001, Residual Heat Removal Pump Overhaul, did not specify that the seal sleeve needed to be installed prior to installing the seal-locking collar.
Additionally, the installation procedure did not specify any post-installation acceptance criteria to ensure the seal is properly seated. An analysis of the seal failure determined that leakage would not exceed the 2 gallon per minute Technical Specification limit but would exceed the 1 gallon per minute administrative limit for emergency core cooling system leakage outside containment.
This finding is more than minor because it was associated with the Barrier Integrity cornerstone attribute of procedural quality and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process," this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since the 2 gallon per minute limit assumed in the post accident dose calculation was preserved and therefore the degraded condition would have no impact on large early release frequency. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810933. This finding did not have a crosscutting aspect since it was not a performance deficiency indicative of current licensee performance.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to terminate refueling water storage tank recirculation results in inadvertent loss of spent fuel pool inventory The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1a, Procedures, for the failure to close Valve BNV0002 during a fill of the spent fuel pool resulting in approximately 2000 gallons of water being inadvertently transferred from the spent fuel pool to the refueling water storage tank. On November 7, 2008, Procedure OTN EC 00001 was performed to add makeup water to the spent fuel pool. Prior to performing the evolution, operations briefed that the refueling water storage tank was on recirculation and that this alignment needed to be secured prior to performing a fill of the spent fuel pool. Following termination of the refueling water storage tank recirculation lineup and after a fill of the spent fuel pool was initiated, the control room received annunciator RWST Lev HILO. The crew recognized that an inadvertent transfer of spent fuel pool water to the refueling water storage tank was in progress and directed that Valves ECV0076 and BNV0002 be closed. It was later discovered that poor communication between operators on the status of Valve BNV0002 resulted in the refueling water storage tank remaining on recirculation during the fill operation.
This finding is more than minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609.04,
 
Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because it only represents a degradation of the radiological barrier function provided by the spent fuel pool. This issue was entered into the licensee's corrective action program as Callaway Action Request 200811692.
This finding had a crosscutting aspect in the area of human performance associated with the work control component because operations personnel failed to effectively communicate work status to the control room.
Inspection Report# : 2008005 (pdf)
Significance:      Sep 24, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Boric Acid Corrosion Control Procedures The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform a corrosion evaluation of boric acid leakage from containment spray Valve ENHV0006 in accordance with Procedure EDP ZZ 01004, Boric Acid Corrosion Control Program. On August 29, 2008, the resident inspectors identified an active packing leak on Valve ENHV0006 with impact to carbon steel components on the valve as evident by discolored, brown boron. The leak, which had been active since February 27, 2007, was caused by a stem imperfection that was previously identified on December 5, 2007. The inspectors noted that Valve ENHV0006 did not have a current boric acid corrosion evaluation despite meeting the screening requirements for an evaluation listed in Procedure EDP ZZ 01004, Boric Acid Corrosion Control Program, Section 4.2. Programmatic boric acid control and work control issues were a key contributor to not recognizing the need for an updated boric acid corrosion evaluation.
This finding is more than minor because, if left uncorrected, the failure to analyze the effects of boric acid corrosion on safety related components could become a more significant safety concern. This finding affected the barrier integrity cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined be of very low safety significance because the finding does not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere, does not represent an actual open pathway in the physical integrity of the reactor containment, and does not involve an actual reduction in function of hydrogen ignitors in the reactor containment. This issue was entered into the licensee's corrective action program as Callaway Action Request 200809351. This finding has a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to interdepartmentally coordinate the impact of changes to the work scope for Valve ENHV0006 such that appropriate personnel could perform the necessary evaluations to assure plant performance.
Inspection Report# : 2008004 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Comply with Radiation Work Permit Requirements The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1.a, which resulted from a failure to comply with radiation work permit instructions. Specifically, on November 2, 2008, during a change out of the chemical and volume control system reactor coolant Filter FBG06, the technicians failed to follow radiation work permit instructions that required notification of the ALARA specialist if the vent port radiation monitor reading was greater than or equal to 1500 millirem per hour to determine if additional briefing requirements were needed. The
 
licensee entered this item into their corrective action program as Callaway Action Request 200811469. As corrective action, the licensee has modified the briefing procedure and modified the radiation work permits to include a requirement to notify radiation protection supervision to evaluate dose rate readings of the vent port and filter housing.
Other corrective actions are being evaluated.
Failure to comply with radiation work permit requirements is a performance deficiency. The finding is greater than minor because it is associated with the cornerstone attribute of exposure control and affected the cornerstone objective, in that, the failure to follow radiation work permit requirements increases the potential for increased dose.
The finding involved workers unplanned, unintended doses or potential of such a dose (resulting from actions or conditions contrary to the radiation work permit). Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, the finding had a crosscutting aspect in the area of human performance, work practices, because the licensee failed to communicate human error prevention techniques during the prejob briefing and ensure that all personnel understood limits stated in the radiation work permit. In addition, personnel proceeded with the filter change out even though radiation levels were significantly higher than anticipated.
Inspection Report# : 2009003 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to comply wity high radiation area entry requirements.
The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.7.1, which resulted from a failure of three individuals to comply with high radiation area entry requirements. Specifically, on October 20, 2008, three engineers touring the reactor building entered a posted high radiation area without signing in on a radiation work permit which allowed entry into a high radiation area, and did not receive a briefing on dose rates in the high radiation area. Shortly after entering the high radiation area, one of the engineers received an electronic dosimeter rate alarm when dose rates in the area exceeded the 50 millirem per hour setpoint. The licensee entered this event into their corrective action program and conducted an Event Review Team meeting to determine the probable causes that led to the event and recommend corrective actions to prevent the event from happening in the future.
Failure to comply with high radiation area entry requirements is a performance deficiency. This finding is greater than minor because it was associated with the cornerstone attribute of exposure control and affected the cornerstone objective, in that, the failure to meet high radiation area entry requirements increases the potential for increased radiation dose. This finding involved an individual workers' unplanned, unintended dose or potential of such dose (resulting from actions or conditions contrary to Technical Specifications) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised.
Additionally, the finding had a crosscutting aspect in the area of human performance, work practices component, because the workers failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2008005 (pdf)
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings
 
pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : August 31, 2009
 
Callaway 3Q/2009 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Response to Feedwater Transient Results in Reactor Trip The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after operator response to an electrical fault on the condensate Pump C motor resulted in an unplanned and unnecessary reactor trip, feedwater isolation, and auxiliary feedwater actuation. On December 11, 2008, Callaway Plant experienced an automatic turbine trip/reactor trip during a power reduction initiated by the operators response to a loss of condensate Pump C. The control room supervisor directed a power reduction without immediately referencing Procedure OTO AE 00001 guidance and without specifying any magnitude or rate limitations on the power reduction.
The balance of plant reactor operator, not aware of the procedural limitations, initiated the power reduction using the turbine controls load limiter potentiometer. This method of turbine load control eliminated all automatic rate-limiting functions. The steam generator levels increased rapidly with sluggish main feedwater regulating valves slowing anticipatory response. The steam generator P-14 high-high level turbine trip/reactor trip occurred about 5 minutes after condensate Pump C had tripped.
This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. The finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively establish clear expectations and standards regarding procedurally directed actions versus actions viewed as necessary to stabilize a plant transient.
Inspection Report# : 2009002 (pdf)
Significance:        Mar 24, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Intermediate Range Hi Flux Reactor Protection System Actuation W hile Shutdown The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after maintenance on intermediate range nuclear Instrument N36 resulted in an unanticipated reactor trip signal and feedwater isolation. On December 12, 2008, Callaway instrumentation and controls maintenance personnel performed work to replace a circuit card associated with the intermediate range nuclear Instrument P 6 bistable. At the time of the maintenance, the plant was in Mode 3 with the reactor trip breakers open. Shortly after beginning work, an intermediate range high flux reactor trip signal was generated. The trip signal was generated because the bypass of the reactor trip bistables is removed upon removal of the control power fuses. With instrument power removed, the solid state protection system perceived a high intermediate range neutron flux condition and generated a reactor trip signal and feedwater isolation. Control room operators responded to the feedwater isolation by starting both motor-driven auxiliary feedwater pumps and restoring steam generator water levels to the program band. The licensee later determined that instrumentation and controls maintenance personnel were unaware that pulling the control power fuses would cause a reactor trip signal and that the step in the work instruction that directed the removal of the control power fuses had not received an adequate review.
This finding was greater than minor because the finding impacted the Initiating Events cornerstone attribute of human
 
performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This issue was entered into the licensee's corrective action program as Callaway Action Request 200812681. The finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to coordinate the impact of changes to the work scope or activity, specifically, the licensee failed to fully evaluate the impact of removal of control power fuses on the work instructions.
Inspection Report# : 2009002 (pdf)
Significance:        Feb 27, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Problems with Fire Protection Impairment Permits An NRC identified violation of License Condition 2.C.(5), Fire Protection, was identified for failing to effectively correct problems with the issuance and establishment of Fire Protection Impairment Permits. After problems were identified in 2006 and 2007, as a corrective action, the licensee conducted training in 2008 on the program requirements in the Maintenance and Operations Departments. Despite this corrective action, the licensee continued to experience failures to request a fire impairment and failures to implement pre-planned impairments. Some failures involved oversight problems for contract workers, who were not addressed in the training. Two procedural violations occurred in late 2008 that involved the failure to establish a Fire Protection Impairment Permit before performing hot work. The licensee has entered the issue into the corrective action program as Callaway Action Request (CAR) 200901638.
The inspectors determined that failing to correct problems associated with the use of required Fire Protection Impairment Permits is a performance deficiency. The finding is more than minor because it affects the protection against external factors attribute of the initiating events cornerstone, and it directly affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using the NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, Phase 1 worksheet, the finding was determined to be of very low safety significance (Green) because the condition represented a low degradation of fire prevention and administrative controls. The cause of the finding is related to the Human Performance cross-cutting component of Work Practices, in that the licensee failed to effectively communicate expectations and personnel failed to follow procedures [H.4.b].
Inspection Report# : 2009006 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to maintain an adequate plant shutdown procedure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after improper isolation of the main steam isolation valves by the Callaway control room operators resulted in a reactor trip signal and auxiliary feedwater actuation on October 11, 2008. Procedure OTG ZZ 00006, "Plant Cooldown Hot Standby to Cold Shutdown," allowed premature main steam isolation valve closures just after entering Mode 4. The operator then decided to reopen main steam isolation Valve A and atmospheric Steam Dump A. This created a significant increase in steam flow from the steam generator which caused the steam generator level to swell up to the P 14 steam generator high level feedwater isolation setpoint. The steam generator levels all decreased to the steam generator narrow range low-low setpoint generating the need for auxiliary feedwater actuation.
This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings," this finding is determined to be of very low safety
 
significance since this finding did not affect the Technical Specification limit for reactor coolant system leakage, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to communicate, in a timely manner, decisions to personnel who have a need to know the information in order to perform work safely.
Inspection Report# : 2008005 (pdf)
Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Failure to evaluate material equivalencies leads to a manual reactor trip The inspectors identified a self-revealing finding for failure of the engineering department to perform a material equivalency evaluation to ensure replacement components do not adversely affect plant operations. On November 11, 2008, Callaway Plant experienced a trip of main feedwater Pump B due to low lube oil pressure. Since the reactor was at greater than 80 percent power, the plant operators inserted a manual reactor trip. Following the reactor trip, maintenance personnel discovered two pieces of o-ring foreign material within main feedwater Pump B bearing oil supply pressure regulating Valve FCV0970. The foreign material was found wrapped around the regulating spring which inhibited valve movement and caused the lube oil low pressure condition. The licensee determined that the ethylene propylene diene M-class type o-ring became pliable when exposed to lube oil and was allowed to fall and be introduced into the system as foreign material. The ethylene propylene diene M-class o-rings had been approved as an equivalent replacement in July 1999 for the vendor recommended Buna-N type o-rings without performing an engineering material equivalency evaluation. Buna-N material is approved for use in petroleum based systems while ethylene propylene diene M-class is not.
This finding is greater than minor because it is associated with the design control attribute of the Initiating Events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using Manual Chapter 0609.04, "Phase 1 -
Initial Screening and Characterization of Findings," the finding is determined to be potentially risk significant because it contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. When evaluated per Manual Chapter 0609 Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," and the Callaway Plant Phase 2 pre-solved table item Failure to Reestablish Main Feedwater, the inspectors determined this finding to be of very low safety significance. This issue was entered into the licensee's corrective action program as Callaway Action Request 200811781. This finding was determined to not have a crosscutting aspect because the performance deficiency is not indicative of current licensee performance.
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance: TBD Sep 30, 2009 Identified By: NRC Item Type: AV Apparent Violation Turbine-driven auxiliary feedwater pump inoperable due to inadequately lubricated trip throttle valve The team identified a self-revealing apparent violation of Technical Specification 3.7.5, Auxiliary Feedwater System, due to the failure to adequately lubricate turbine-driven auxiliary feedwater pump trip throttle valve FCHV0312. During May 25, 2009, surveillance testing, the turbine-driven auxiliary feedwater pump did not start as expected due to hardened grease on the valve spindle of FCHV0312. The previous lubrication preventative maintenance had been missed and lack of lubrication increased friction between the sliding nut and spindle preventing FCHV0312 from opening. Following lubrication FCHV0312 and the turbine-driven auxiliary feedwater pump tested satisfactorily. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200904216.
This finding is greater than minor because it was associated with the equipment performance attribute of the
 
Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as potentially risk significant since the finding represented a loss of system safety function because the turbine-driven auxiliary feedwater pump PAL02 failing eliminates the capability of the plant to cope with a station blackout. The finding required a Phase 2 analysis. When evaluated per Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and the Callaway Plant Phase 2 pre-solved table item Turbine Driven Auxiliary Feedwater Pump Fails to Start, the inspectors determined this finding to be potentially risk significant. The finding was forwarded to a senior reactor analyst for review. The preliminary outcome of the Phase 3 significance determination analysis, Attachment 4, determined the finding was of low to moderate safety significance.
The inspectors determined that this finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to follow the procedural guidance provided when changing the scope of a preventive maintenance task.
Inspection Report# : 2009009 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Lubrication Procedure for Valve FCHV0312 The team identified a noncited violation of Technical Specification 5.4.1.a, Procedures, for the failure to provide adequate procedural guidance for the lubrication of auxiliary feedwater pump turbine trip throttle valve FCHV0312.
The inspectors found that 2002 corrective actions to improve the lubrication procedure were not fully developed and the procedure lubrication guidance was ambiguous in that it did not specify the amount of lubricant to apply or what valve subcomponents to lubricate. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905032.
This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not have a crosscutting aspect since the 2003 lubrication procedure revision was not reflective of current licensee performance.
Inspection Report# : 2009009 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately evaluate the use of Mobile 28 grease for the turbine-driven auxiliary feedwater pump trip throttle valve.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to adequately evaluate the use of Mobile 28 grease for the turbine-driven auxiliary feedwater pump trip throttle valve. The licensees 1995 evaluation included no documentation for the appropriate relubrication interval of the valve. Additionally, the inspectors identified that the valve exhibited temperatures ranging from 235°F to near 300°F compared to the 215°F valve temperature used in the evaluation. The inspectors questioned if the use of Mobile 28 grease was appropriate since operating experience suggests that Mobile 28 grease has a tendency to thicken and harden at temperatures exceeding 250°F and elevated temperatures increased the lubricants tendency to lose oils and could result in increased stem friction. Following questioning by the inspectors, the licensee initiated Callaway Action Request 200905067 and Request for Resolution 200905651 to determine if Mobile 28 grease was an appropriate lubricant for valve FCHV0312.
 
This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not to have a crosscutting aspect since the inadequate 1995 lubrication evaluation was not reflective of current licensee performance.
Inspection Report# : 2009009 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to enter conditions adverse to quality associated with the turbine-driven auxiliary feedwater pump trip throttle valve into the corrective action program.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Callaway Procedure APA ZZ 00500, Corrective Action Program. Specifically, licensee personnel failed to initiate Callaway action requests for adverse conditions of high hand wheel forces, galled subcomponents, and hardened, gritty grease found during the 2007 rebuild of the spare turbine-driven auxiliary feedwater pump trip throttle valve FCHV0312. The licensee has entered this issue into their corrective action program as Callaway Action Request 200905053.
This finding is greater than minor because, if left uncorrected, failure to fully utilize the corrective action program could become a more significant safety concern. The inspectors determined that this finding impacted the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The cause of this finding is related to the problem identification and resolution crosscutting component of the corrective action program because licensee personnel failed to implement a corrective action program with a low threshold for identifying issues.
Inspection Report# : 2009009 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure turbine-driven auxiliary feedwater pump is operable prior to entry into Mode 3.
The team identified a noncited violation of Technical Specification Limiting Condition for Operation 3.0.4 for entering Mode 3 with the turbine-driven auxiliary feedwater pump inoperable. Specifically, on November 3, 2008, while in Mode 4 for Refueling Outage 16, an unexpected overspeed trip of the turbine occurred during postmaintenance testing. Callaway operations staff inappropriately concluded that a water slug from the auxiliary steam line was the cause of the turbine overspeed. Following entry into Mode 3, during preparations for turbine-driven auxiliary feedwater pump testing, the licensee found the servo control valve installed during the outage was faulty. When questioned by the inspectors, the licensee determined that the faulty servo control valve discovered in Mode 3 was responsible for the overspeed of the turbine-driven auxiliary feedwater pump that occurred in Mode 4 and that the equipment was inoperable during the mode change that occurred on November 4, 2008. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905313.
This finding is greater than minor because it is associated with the Mitigating Systems Cornerstone attribute of equipment performance and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04,
 
Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The inspectors determined that this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to fully evaluate the overspeed of the turbine-driven auxiliary feedwater pump that occurred on November 3, 2008.
Inspection Report# : 2009009 (pdf)
Significance:      Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions for Essential Service Water Pump Cable Underground Electrical Vault Seals The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
associated with the licensees failure to take prompt corrective actions prevent continuous submergence of essential service water pump kerite insulated power cables. The continuously submerged environment for these cables existed because the two vaults containing these cables (MH-01N and MH-01S) had inadequate seals needed to protect the vaults from incoming surface water. Callaway Action Request 200201916 stated that all medium voltage cables of concern were located more than 4 feet above the basemat of the vault and thus were not in a submerged condition. The Callaway action request noted that the seals at the top lid were the source of water intrusion and that the seal design was inadequate. On July 9 and 22, 2009, the resident inspectors, along with Callaway plant engineers, inspected the two essential service water underground vaults. The north vault (train A) was found to have water covering the two safety related upper cable trays. Contrary to the Callaway Action Request 200201916 evaluation, the cable trays were about 2.5 feet and 3.5 feet from the basemat of the vault floor. During these 2009 inspections it was noted that the same lid design deficiency identified in Callaway Action Request 200201916 still existed. This led to the discovery that the Callaway Action Request actions from 2002 had not been completely performed. The only significant corrective action had been to increase the inspection frequency to once every three years.
The licensee has subsequently taken measures to improve the seals and written Callaway Action Request/Request for Resolution 200905838 to further evaluate this issue. This finding is more than minor because it affected the Mitigating Systems Cornerstone attribute of design control for ensuring the availability, reliability, and capability of safety systems. Using Manual Chapter 0609.04, Phase 1 - Initial screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded seals were a design or qualification deficiency confirmed not to result in loss of operability. The inspectors determined that the finding has no crosscutting aspect as the performance deficiencies were not reflective of current performance. The licensee entered this item into their corrective action program as Callaway Action Request 200908855.
Inspection Report# : 2009004 (pdf)
Significance: SL-IV Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correctly Identify Safety System Functional Failures in a Licensee Event Report The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for a failure to report two examples of safety system functional failures in licensee event reports within 60 days after discovery of events requiring a report. The two examples were:
* March 26, 2008, discovery that operation of containment air coolers in fast speed, during a period of higher than normal containment pressure, could open the air coolers fast speed thermal overload device rendering all the coolers incapable of automatically restarting in slow speed
* May 21, 2008, discovery of a 6.6 cubic foot void of air in the common suction piping capable of affecting the function of both of the safety injection system pumps For each example, the inspectors reviewed the licensees reportability evaluation and associated past operability
 
reviews and determined each event was reportable per 10 CFR 50.73(a)(2)(v) since each example resulted in a condition which affected both trains of a system described in the Final Safety Analysis Report that was needed to mitigate the consequences of an accident. Alternate safety systems accident mitigation is not permitted as a reason to not report the discovery of the conditions. The licensee also failed to report these failures to the NRC performance indicator database because of the failure to include the safety system functional failure in each respective licensee event report.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. The licensee planned to update the associated license event reports as described in Callaway Action Request 200904980. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure, through adequate training, that its staff understood the guidance documents pertaining to the 10 CFR 50.73 rule.
Inspection Report# : 2009004 (pdf)
Significance: SL-IV Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit Complete and Accurate Risk Information for a Requested License Amendment The inspectors identified a noncited violation of 10 CFR 50.9, "Completeness and Accuracy of Information," when AmerenUE failed to submit complete and accurate quantification of risk contributors associated with a license amendment supporting a modification to replace the underground portion of the essential service water system Train B piping with high density polyethelene pipe. The inspectors questioned the risk impact of a possible control room fire which led to the discovery that the licensee had not followed their process for screening out fire areas. The licensee entered this item into their corrective action program as Callaway Action Request 200902810 and also submitted an update to License Amendment 191 to correctly account for the control room fire risk.
This finding affects the Mitigating Systems cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect because the licensees failure to thoroughly review and submit the risk for control room fires was not part of a corrective action process, but instead an oversight by the licensing review process.
Inspection Report# : 2009003 (pdf)
Significance:      Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Controls of Crane Work Above the Protected Train of Essential Service Water The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) associated with the licensees failure to adequately assess and manage risk associated with crane work over the essential service water system Train A. On March 31, 2009, the licensee performed work in the vicinity of the protected essential service water system train which included movement of 1800 pound sand bags over the protected train piping. After questioning by the resident inspectors, the licensee determined that the lifts were not conducted in accordance with station procedures since the requirements of a required engineering judgment memo were not translated into work documents. The licensee entered this item into their corrective action program as Callaway Action Request 200902726.
The finding affected the Mitigating Systems cornerstone and was determined to be more than minor because the licensee failed to implement the prescribed significant compensatory measures associated with crane work in the vicinity of safe shutdown equipment. This finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to include appropriate risk insights in planned work activities.
Inspection Report# : 2009003 (pdf)
 
Significance:      Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate, At Power, Risk Assessment for Maintenance Activities on One Train of Essential Service Water and Emergency Diesel Generator The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) associated with the licensees failure to perform an adequate risk assessment for planned maintenance on the emergency diesel generator Train A and essential service water pump Train A. On April 28, 2009, Callaway Plant operators removed the emergency diesel generator Train A and essential service water pump Train A from service. The inspectors' review of the plant risk profile for the in-progress maintenance activity uncovered that this risk had not been accounted for by the plant safety monitor tool.
The licensee entered this item into their corrective action program as Callaway Action Request 200903480 The finding is more than minor because the risk, when correctly assessed, put the plant into a higher risk category for large early release frequency. Also the licensee risk assessment failed to consider risk significant systems, structures, and components and support systems that were unavailable during the maintenance. This finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to appropriately plan work activities consistent with nuclear safety by incorporating risk insights.
Inspection Report# : 2009003 (pdf)
Significance:      Apr 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Suitable Replacement Parts Essential for Emergency Diesel Generator Train B The inspectors identifed a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" after the licensee failed to adequately select suitable replacement parts essential to the operation of emergency diesel generator Train B. On December 24, 2008, during performance of Procedure OSP-NE-0001B, "Standby Diesel Generator B Periodic Tests, " Callaway operations personnel identified that the emergency diesel generator Train B had an approximately 0.82 gallon per minute jacket water leak resulting in operators declaring the equipment inoperable.
Upon removal, the gasket was found to be soft and extruding from the flange edge. The licensee originally concluded the gasket failed due to vibrations associated with engine shutdown but altered that conclusion after discussions witht he resident inspectors and additional investigation. The licensee ultimately determined that the cause of the failure was due to incorrect gasket material being used during Job W200773 performed on October 16, 1999. The gasket was 1/8" thick which resulted ina lack of compression. Since the gaskets are composed of an aramid fiberous material, the lack of compression allowed the gasket to absorb water and soften. The leak identified on December 24, 2008, developed once the gasket softened sufficiently to extrude from the flange edge. This issue has been entered into the licensee's corrective action program as Callaway Action Request 200812985.
This finding was greater than minor because it was associated with the mitigating systems cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings, " this finding was determined to represent an actual loss of safety function of a single train for greater thatn its Technical Specification allowed outage time. When evaluated per Manual Chapter 0609 Appendiz A, "Determining the Significance of Reactor Inspection Finding for At-Power Situations," and the Callaway Plant P hae 2 pre-solved table item "Diesel Generator Fails to Run after Start," the inspectors determined this finding to be potentially risk significant. This finding was forwarded to a senior reactor analyst for review. The results of the senior reactor analyst's Phase 3 analysis determined the finding to be of very low safety significance. This finding did not have a crosscutting aspect since it was not a performance deficiency indicative of current licensee performance.
Inspection Report# : 2009007 (pdf)
Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation
 
Inadequate shutdown risk assessment for maintenance activities in the reactor building.
The inspectors identified a noncited violation of 10 CFR 50.65(a)(4), for failure to adequately assess and manage shutdown risk associated with maintenance activities in the reactor building. Specifically, on October 15, 2008, the inspectors found foreign material exclusion covers installed on the Train B containment recirculation sump. The covers were installed on October 14, 2008, per the direction of the containment coordinator without notification to the control room. The covers were installed to prevent debris from entering the sump. Following discussions with operations personnel, the inspectors found that the Train B containment recirculation sump was inappropriately credited in the licensees shutdown safety assessment. An updated shutdown safety assessment was performed and it was determined that plant risk remained yellow.
This finding is greater than minor because the licensees risk assessment inappropriately credited risk-significant structures, systems and components that were unavailable during maintenance. This finding affected the Mitigating Systems cornerstone. Using Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the finding was found to be of very low safety significance because the licensee maintained two trains of decay heat removal operable and adequate equipment was available to support feed and bleed operations for at least 24 hours. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810540.
This finding had a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to obtain interdisciplinary input on safety-significant or risk-significant decisions. Specifically, the containment coordinator made a decision affecting the availability of the containment recirculation sumps without consulting the control room to determine the impact on plant risk.
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to ensure the suitability of the design of the resideual heat removal Train A pump room cooler The inspectors identified a self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after a trip of the residual heat removal Train A room cooler fan revealed that AmerenUE had not adequately selected and reviewed the suitability of the newly installed fan motor thermal overloads. Additionally, the NRC inspectors identified that the postmaintenance testing prescribed for the modified fan motor breaker did not allow sufficient time to challenge the thermal overload settings. On October 8, 2008, residual heat removal Train A room cooler fan shut down after only 22 minutes of run time. The breaker replacement modification used a calculation originally performed for the initial design of the old breaker which did not account for the cooler fan motor being a 20 horsepower motor nameplated down to a 10 horsepower rating.
This finding is greater than minor because it is similar to Manual Chapter 0612 "Examples of Minor Issues," Example 3j, in that the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of the component and a significant programmatic deficiency associated with postmaintenance test requirements was identified that could lead to worse errors if uncorrected. The inspectors determined that the finding impacted the Mitigating Systems cornerstone. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than Technical Specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810223. The inspectors determined that this finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action component because the AmerenUE modification for certain motor control center breakers failed to have a low enough threshold to identify fan motor rating and thermal overload setting errors.
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation
 
Failure to adequately implement plant equipment control tagout procedure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after improper restoration of the essential service water supply to the emergency diesel generator Train A lubricating oil cooler resulted in significant water flow into the emergency diesel room on October 22, 2008. Two restoration evolutions associated with the essential service water and the emergency diesel generator systems had been proceeding in parallel. The reactor operator restoring the emergency diesel generator assumed the essential service water supply was to remain isolated to the emergency diesel generator and thus changed the already approved worker protection assurance Clearance 71899 to leave the oil cooler drain valve open with no tag. Starting the essential service water pump pressurized the drain valve and produced significant water spray flow into the emergency diesel generator room until noticed by a diesel vendor representative about 30 minutes later.
This finding was greater than minor because if left uncorrected the deficiencies could become a more significant safety concern. The finding affected the Mitigating Systems cornerstone. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings," this finding is determined to be of very low safety significance since this finding was not a design or qualification deficiency, did not represent a loss of system or train safety function and did not screen as potentially risk significant due to a flooding initiating event using the criteria on the characterization worksheet. This finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee's pre-job briefing, self- and peer-checking, and proper documentation of activity were inadequate to overcome worker protection assurance clearance process problems and an inexperienced operating supervisor. These less than adequate worker practices resulted in personnel proceeding in the face of uncertainty.
Inspection Report# : 2008005 (pdf)
Barrier Integrity Significance:      Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate maintenance procedure resulted in residual heat removal mechanical seal failure The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1a, Procedures, for inadequate procedural guidance that resulted in the failure of the residual heat removal Train A pump mechanical seal.
On October 22, 2008, the licensee discovered a solid stream of water issuing from the residual heat removal Train A pump mechanical seal. The failure occurred because of installation difficulties encountered on October 8, 2008, when the seal sleeve was installed with the seal locking collar engaged. This configuration resulted in increased loading on the seal seating surfaces that resulted in surface chipping and led to seal failure after approximately 48 hours of shutdown cooling operation. Mechanical seal replacement Procedure MPM EJ QP001, Residual Heat Removal Pump Overhaul, did not specify that the seal sleeve needed to be installed prior to installing the seal-locking collar.
Additionally, the installation procedure did not specify any post-installation acceptance criteria to ensure the seal is properly seated. An analysis of the seal failure determined that leakage would not exceed the 2 gallon per minute Technical Specification limit but would exceed the 1 gallon per minute administrative limit for emergency core cooling system leakage outside containment.
This finding is more than minor because it was associated with the Barrier Integrity cornerstone attribute of procedural quality and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process," this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since the 2 gallon per minute limit assumed in the post accident dose calculation was preserved and therefore the degraded condition would have no impact on large early release frequency. This issue was entered into the licensee's corrective action program as Callaway Action Request 200810933. This finding did not have a crosscutting aspect since it was not a performance deficiency indicative of current licensee performance.
 
Inspection Report# : 2008005 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to terminate refueling water storage tank recirculation results in inadvertent loss of spent fuel pool inventory The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1a, Procedures, for the failure to close Valve BNV0002 during a fill of the spent fuel pool resulting in approximately 2000 gallons of water being inadvertently transferred from the spent fuel pool to the refueling water storage tank. On November 7, 2008, Procedure OTN EC 00001 was performed to add makeup water to the spent fuel pool. Prior to performing the evolution, operations briefed that the refueling water storage tank was on recirculation and that this alignment needed to be secured prior to performing a fill of the spent fuel pool. Following termination of the refueling water storage tank recirculation lineup and after a fill of the spent fuel pool was initiated, the control room received annunciator RWST Lev HILO. The crew recognized that an inadvertent transfer of spent fuel pool water to the refueling water storage tank was in progress and directed that Valves ECV0076 and BNV0002 be closed. It was later discovered that poor communication between operators on the status of Valve BNV0002 resulted in the refueling water storage tank remaining on recirculation during the fill operation.
This finding is more than minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or releases. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because it only represents a degradation of the radiological barrier function provided by the spent fuel pool. This issue was entered into the licensee's corrective action program as Callaway Action Request 200811692.
This finding had a crosscutting aspect in the area of human performance associated with the work control component because operations personnel failed to effectively communicate work status to the control room.
Inspection Report# : 2008005 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Comply with Radiation Work Permit Requirements The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1.a, which resulted from a failure to comply with radiation work permit instructions. Specifically, on November 2, 2008, during a change out of the chemical and volume control system reactor coolant Filter FBG06, the technicians failed to follow radiation work permit instructions that required notification of the ALARA specialist if the vent port radiation monitor reading was greater than or equal to 1500 millirem per hour to determine if additional briefing requirements were needed. The licensee entered this item into their corrective action program as Callaway Action Request 200811469. As corrective action, the licensee has modified the briefing procedure and modified the radiation work permits to include a requirement to notify radiation protection supervision to evaluate dose rate readings of the vent port and filter housing.
Other corrective actions are being evaluated.
Failure to comply with radiation work permit requirements is a performance deficiency. The finding is greater than minor because it is associated with the cornerstone attribute of exposure control and affected the cornerstone objective, in that, the failure to follow radiation work permit requirements increases the potential for increased dose.
The finding involved workers unplanned, unintended doses or potential of such a dose (resulting from actions or
 
conditions contrary to the radiation work permit). Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, the finding had a crosscutting aspect in the area of human performance, work practices, because the licensee failed to communicate human error prevention techniques during the prejob briefing and ensure that all personnel understood limits stated in the radiation work permit. In addition, personnel proceeded with the filter change out even though radiation levels were significantly higher than anticipated.
Inspection Report# : 2009003 (pdf)
Significance:        Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to comply wity high radiation area entry requirements.
The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.7.1, which resulted from a failure of three individuals to comply with high radiation area entry requirements. Specifically, on October 20, 2008, three engineers touring the reactor building entered a posted high radiation area without signing in on a radiation work permit which allowed entry into a high radiation area, and did not receive a briefing on dose rates in the high radiation area. Shortly after entering the high radiation area, one of the engineers received an electronic dosimeter rate alarm when dose rates in the area exceeded the 50 millirem per hour setpoint. The licensee entered this event into their corrective action program and conducted an Event Review Team meeting to determine the probable causes that led to the event and recommend corrective actions to prevent the event from happening in the future.
Failure to comply with high radiation area entry requirements is a performance deficiency. This finding is greater than minor because it was associated with the cornerstone attribute of exposure control and affected the cornerstone objective, in that, the failure to meet high radiation area entry requirements increases the potential for increased radiation dose. This finding involved an individual workers' unplanned, unintended dose or potential of such dose (resulting from actions or conditions contrary to Technical Specifications) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised.
Additionally, the finding had a crosscutting aspect in the area of human performance, work practices component, because the workers failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2008005 (pdf)
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : December 10, 2009
 
Callaway 4Q/2009 Plant Inspection Findings Initiating Events Significance:        Mar 24, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Response to Feedwater Transient Results in Reactor Trip The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after operator response to an electrical fault on the condensate Pump C motor resulted in an unplanned and unnecessary reactor trip, feedwater isolation, and auxiliary feedwater actuation. On December 11, 2008, Callaway Plant experienced an automatic turbine trip/reactor trip during a power reduction initiated by the operators response to a loss of condensate Pump C. The control room supervisor directed a power reduction without immediately referencing Procedure OTO AE 00001 guidance and without specifying any magnitude or rate limitations on the power reduction.
The balance of plant reactor operator, not aware of the procedural limitations, initiated the power reduction using the turbine controls load limiter potentiometer. This method of turbine load control eliminated all automatic rate-limiting functions. The steam generator levels increased rapidly with sluggish main feedwater regulating valves slowing anticipatory response. The steam generator P-14 high-high level turbine trip/reactor trip occurred about 5 minutes after condensate Pump C had tripped.
This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. The finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively establish clear expectations and standards regarding procedurally directed actions versus actions viewed as necessary to stabilize a plant transient.
Inspection Report# : 2009002 (pdf)
Significance:        Mar 24, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Intermediate Range Hi Flux Reactor Protection System Actuation W hile Shutdown The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after maintenance on intermediate range nuclear Instrument N36 resulted in an unanticipated reactor trip signal and feedwater isolation. On December 12, 2008, Callaway instrumentation and controls maintenance personnel performed work to replace a circuit card associated with the intermediate range nuclear Instrument P 6 bistable. At the time of the maintenance, the plant was in Mode 3 with the reactor trip breakers open. Shortly after beginning work, an intermediate range high flux reactor trip signal was generated. The trip signal was generated because the bypass of the reactor trip bistables is removed upon removal of the control power fuses. With instrument power removed, the solid state protection system perceived a high intermediate range neutron flux condition and generated a reactor trip signal and feedwater isolation. Control room operators responded to the feedwater isolation by starting both motor-driven auxiliary feedwater pumps and restoring steam generator water levels to the program band. The licensee later determined that instrumentation and controls maintenance personnel were unaware that pulling the control power fuses would cause a reactor trip signal and that the step in the work instruction that directed the removal of the control power fuses had not received an adequate review.
This finding was greater than minor because the finding impacted the Initiating Events cornerstone attribute of human
 
performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This issue was entered into the licensee's corrective action program as Callaway Action Request 200812681. The finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to coordinate the impact of changes to the work scope or activity, specifically, the licensee failed to fully evaluate the impact of removal of control power fuses on the work instructions.
Inspection Report# : 2009002 (pdf)
Significance:        Feb 27, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Problems with Fire Protection Impairment Permits An NRC identified violation of License Condition 2.C.(5), Fire Protection, was identified for failing to effectively correct problems with the issuance and establishment of Fire Protection Impairment Permits. After problems were identified in 2006 and 2007, as a corrective action, the licensee conducted training in 2008 on the program requirements in the Maintenance and Operations Departments. Despite this corrective action, the licensee continued to experience failures to request a fire impairment and failures to implement pre-planned impairments. Some failures involved oversight problems for contract workers, who were not addressed in the training. Two procedural violations occurred in late 2008 that involved the failure to establish a Fire Protection Impairment Permit before performing hot work. The licensee has entered the issue into the corrective action program as Callaway Action Request (CAR) 200901638.
The inspectors determined that failing to correct problems associated with the use of required Fire Protection Impairment Permits is a performance deficiency. The finding is more than minor because it affects the protection against external factors attribute of the initiating events cornerstone, and it directly affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using the NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, Phase 1 worksheet, the finding was determined to be of very low safety significance (Green) because the condition represented a low degradation of fire prevention and administrative controls. The cause of the finding is related to the Human Performance cross-cutting component of Work Practices, in that the licensee failed to effectively communicate expectations and personnel failed to follow procedures [H.4.b].
Inspection Report# : 2009006 (pdf)
Mitigating Systems Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Turbine-driven auxiliary feedwater pump inoperable due to inadequately lubricated trip throttle valve The team identified a self-revealing apparent violation of Technical Specification 3.7.5, Auxiliary Feedwater System, due to the failure to adequately lubricate turbine-driven auxiliary feedwater pump trip throttle valve FCHV0312. During May 25, 2009, surveillance testing, the turbine-driven auxiliary feedwater pump did not start as expected due to hardened grease on the valve spindle of FCHV0312. The previous lubrication preventative maintenance had been missed and lack of lubrication increased friction between the sliding nut and spindle preventing FCHV0312 from opening. Following lubrication FCHV0312 and the turbine-driven auxiliary feedwater pump tested satisfactorily. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200904216.
 
This finding is greater than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as potentially risk significant since the finding represented a loss of system safety function because the turbine-driven auxiliary feedwater pump PAL02 failing eliminates the capability of the plant to cope with a station blackout. The finding required a Phase 2 analysis. When evaluated per Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and the Callaway Plant Phase 2 pre-solved table item Turbine Driven Auxiliary Feedwater Pump Fails to Start, the inspectors determined this finding to be potentially risk significant. The finding was forwarded to a senior reactor analyst for review. The preliminary outcome of the Phase 3 significance determination analysis, Attachment 4, determined the finding was of low to moderate safety significance.
The inspectors determined that this finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to follow the procedural guidance provided when changing the scope of a preventive maintenance task.
Inspection Report# : 2009009 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Lubrication Procedure for Valve FCHV0312 The team identified a noncited violation of Technical Specification 5.4.1.a, Procedures, for the failure to provide adequate procedural guidance for the lubrication of auxiliary feedwater pump turbine trip throttle valve FCHV0312.
The inspectors found that 2002 corrective actions to improve the lubrication procedure were not fully developed and the procedure lubrication guidance was ambiguous in that it did not specify the amount of lubricant to apply or what valve subcomponents to lubricate. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905032.
This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not have a crosscutting aspect since the 2003 lubrication procedure revision was not reflective of current licensee performance.
Inspection Report# : 2009009 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately evaluate the use of Mobile 28 grease for the turbine-driven auxiliary feedwater pump trip throttle valve.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to adequately evaluate the use of Mobile 28 grease for the turbine-driven auxiliary feedwater pump trip throttle valve. The licensees 1995 evaluation included no documentation for the appropriate relubrication interval of the valve. Additionally, the inspectors identified that the valve exhibited temperatures ranging from 235°F to near 300°F compared to the 215°F valve temperature used in the evaluation. The inspectors questioned if the use of Mobile 28 grease was appropriate since operating experience suggests that Mobile 28 grease has a tendency to thicken and harden at temperatures exceeding 250°F and elevated temperatures increased the lubricants tendency to lose oils and could result in increased stem friction. Following questioning by the inspectors, the licensee initiated Callaway Action Request 200905067 and Request for Resolution 200905651 to determine if Mobile 28 grease was an appropriate lubricant for valve FCHV0312.
 
This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not to have a crosscutting aspect since the inadequate 1995 lubrication evaluation was not reflective of current licensee performance.
Inspection Report# : 2009009 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to enter conditions adverse to quality associated with the turbine-driven auxiliary feedwater pump trip throttle valve into the corrective action program.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Callaway Procedure APA ZZ 00500, Corrective Action Program. Specifically, licensee personnel failed to initiate Callaway action requests for adverse conditions of high hand wheel forces, galled subcomponents, and hardened, gritty grease found during the 2007 rebuild of the spare turbine-driven auxiliary feedwater pump trip throttle valve FCHV0312. The licensee has entered this issue into their corrective action program as Callaway Action Request 200905053.
This finding is greater than minor because, if left uncorrected, failure to fully utilize the corrective action program could become a more significant safety concern. The inspectors determined that this finding impacted the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The cause of this finding is related to the problem identification and resolution crosscutting component of the corrective action program because licensee personnel failed to implement a corrective action program with a low threshold for identifying issues.
Inspection Report# : 2009009 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure turbine-driven auxiliary feedwater pump is operable prior to entry into Mode 3.
The team identified a noncited violation of Technical Specification Limiting Condition for Operation 3.0.4 for entering Mode 3 with the turbine-driven auxiliary feedwater pump inoperable. Specifically, on November 3, 2008, while in Mode 4 for Refueling Outage 16, an unexpected overspeed trip of the turbine occurred during postmaintenance testing. Callaway operations staff inappropriately concluded that a water slug from the auxiliary steam line was the cause of the turbine overspeed. Following entry into Mode 3, during preparations for turbine-driven auxiliary feedwater pump testing, the licensee found the servo control valve installed during the outage was faulty. When questioned by the inspectors, the licensee determined that the faulty servo control valve discovered in Mode 3 was responsible for the overspeed of the turbine-driven auxiliary feedwater pump that occurred in Mode 4 and that the equipment was inoperable during the mode change that occurred on November 4, 2008. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905313.
This finding is greater than minor because it is associated with the Mitigating Systems Cornerstone attribute of equipment performance and it affected the cornerstone objective to ensure the availability, reliability, and capability of
 
systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The inspectors determined that this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to fully evaluate the overspeed of the turbine-driven auxiliary feedwater pump that occurred on November 3, 2008.
Inspection Report# : 2009009 (pdf)
Significance:      Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions for Essential Service Water Pump Cable Underground Electrical Vault Seals The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
associated with the licensees failure to take prompt corrective actions prevent continuous submergence of essential service water pump kerite insulated power cables. The continuously submerged environment for these cables existed because the two vaults containing these cables (MH-01N and MH-01S) had inadequate seals needed to protect the vaults from incoming surface water. Callaway Action Request 200201916 stated that all medium voltage cables of concern were located more than 4 feet above the basemat of the vault and thus were not in a submerged condition. The Callaway action request noted that the seals at the top lid were the source of water intrusion and that the seal design was inadequate. On July 9 and 22, 2009, the resident inspectors, along with Callaway plant engineers, inspected the two essential service water underground vaults. The north vault (train A) was found to have water covering the two safety related upper cable trays. Contrary to the Callaway Action Request 200201916 evaluation, the cable trays were about 2.5 feet and 3.5 feet from the basemat of the vault floor. During these 2009 inspections it was noted that the same lid design deficiency identified in Callaway Action Request 200201916 still existed. This led to the discovery that the Callaway Action Request actions from 2002 had not been completely performed. The only significant corrective action had been to increase the inspection frequency to once every three years.
The licensee has subsequently taken measures to improve the seals and written Callaway Action Request/Request for Resolution 200905838 to further evaluate this issue. This finding is more than minor because it affected the Mitigating Systems Cornerstone attribute of design control for ensuring the availability, reliability, and capability of safety systems. Using Manual Chapter 0609.04, Phase 1 - Initial screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded seals were a design or qualification deficiency confirmed not to result in loss of operability. The inspectors determined that the finding has no crosscutting aspect as the performance deficiencies were not reflective of current performance. The licensee entered this item into their corrective action program as Callaway Action Request 200908855.
Inspection Report# : 2009004 (pdf)
Significance: SL-IV Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correctly Identify Safety System Functional Failures in a Licensee Event Report The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for a failure to report two examples of safety system functional failures in licensee event reports within 60 days after discovery of events requiring a report. The two examples were:
* March 26, 2008, discovery that operation of containment air coolers in fast speed, during a period of higher than normal containment pressure, could open the air coolers fast speed thermal overload device rendering all the coolers incapable of automatically restarting in slow speed
* May 21, 2008, discovery of a 6.6 cubic foot void of air in the common suction piping capable of affecting the function of both of the safety injection system pumps
 
For each example, the inspectors reviewed the licensees reportability evaluation and associated past operability reviews and determined each event was reportable per 10 CFR 50.73(a)(2)(v) since each example resulted in a condition which affected both trains of a system described in the Final Safety Analysis Report that was needed to mitigate the consequences of an accident. Alternate safety systems accident mitigation is not permitted as a reason to not report the discovery of the conditions. The licensee also failed to report these failures to the NRC performance indicator database because of the failure to include the safety system functional failure in each respective licensee event report.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. The licensee planned to update the associated license event reports as described in Callaway Action Request 200904980. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure, through adequate training, that its staff understood the guidance documents pertaining to the 10 CFR 50.73 rule.
Inspection Report# : 2009004 (pdf)
Significance: SL-IV Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit Complete and Accurate Risk Information for a Requested License Amendment The inspectors identified a noncited violation of 10 CFR 50.9, "Completeness and Accuracy of Information," when AmerenUE failed to submit complete and accurate quantification of risk contributors associated with a license amendment supporting a modification to replace the underground portion of the essential service water system Train B piping with high density polyethelene pipe. The inspectors questioned the risk impact of a possible control room fire which led to the discovery that the licensee had not followed their process for screening out fire areas. The licensee entered this item into their corrective action program as Callaway Action Request 200902810 and also submitted an update to License Amendment 191 to correctly account for the control room fire risk.
This finding affects the Mitigating Systems cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect because the licensees failure to thoroughly review and submit the risk for control room fires was not part of a corrective action process, but instead an oversight by the licensing review process.
Inspection Report# : 2009003 (pdf)
Significance:      Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Controls of Crane Work Above the Protected Train of Essential Service Water The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) associated with the licensees failure to adequately assess and manage risk associated with crane work over the essential service water system Train A. On March 31, 2009, the licensee performed work in the vicinity of the protected essential service water system train which included movement of 1800 pound sand bags over the protected train piping. After questioning by the resident inspectors, the licensee determined that the lifts were not conducted in accordance with station procedures since the requirements of a required engineering judgment memo were not translated into work documents. The licensee entered this item into their corrective action program as Callaway Action Request 200902726.
The finding affected the Mitigating Systems cornerstone and was determined to be more than minor because the licensee failed to implement the prescribed significant compensatory measures associated with crane work in the vicinity of safe shutdown equipment. This finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to include appropriate risk insights in planned work activities.
 
Inspection Report# : 2009003 (pdf)
Significance:      Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate, At Power, Risk Assessment for Maintenance Activities on One Train of Essential Service Water and Emergency Diesel Generator The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) associated with the licensees failure to perform an adequate risk assessment for planned maintenance on the emergency diesel generator Train A and essential service water pump Train A. On April 28, 2009, Callaway Plant operators removed the emergency diesel generator Train A and essential service water pump Train A from service. The inspectors' review of the plant risk profile for the in-progress maintenance activity uncovered that this risk had not been accounted for by the plant safety monitor tool.
The licensee entered this item into their corrective action program as Callaway Action Request 200903480 The finding is more than minor because the risk, when correctly assessed, put the plant into a higher risk category for large early release frequency. Also the licensee risk assessment failed to consider risk significant systems, structures, and components and support systems that were unavailable during the maintenance. This finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to appropriately plan work activities consistent with nuclear safety by incorporating risk insights.
Inspection Report# : 2009003 (pdf)
Significance:      Apr 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Suitable Replacement Parts Essential for Emergency Diesel Generator Train B The inspectors identifed a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" after the licensee failed to adequately select suitable replacement parts essential to the operation of emergency diesel generator Train B. On December 24, 2008, during performance of Procedure OSP-NE-0001B, "Standby Diesel Generator B Periodic Tests, " Callaway operations personnel identified that the emergency diesel generator Train B had an approximately 0.82 gallon per minute jacket water leak resulting in operators declaring the equipment inoperable.
Upon removal, the gasket was found to be soft and extruding from the flange edge. The licensee originally concluded the gasket failed due to vibrations associated with engine shutdown but altered that conclusion after discussions witht he resident inspectors and additional investigation. The licensee ultimately determined that the cause of the failure was due to incorrect gasket material being used during Job W200773 performed on October 16, 1999. The gasket was 1/8" thick which resulted ina lack of compression. Since the gaskets are composed of an aramid fiberous material, the lack of compression allowed the gasket to absorb water and soften. The leak identified on December 24, 2008, developed once the gasket softened sufficiently to extrude from the flange edge. This issue has been entered into the licensee's corrective action program as Callaway Action Request 200812985.
This finding was greater than minor because it was associated with the mitigating systems cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings, " this finding was determined to represent an actual loss of safety function of a single train for greater thatn its Technical Specification allowed outage time. When evaluated per Manual Chapter 0609 Appendiz A, "Determining the Significance of Reactor Inspection Finding for At-Power Situations," and the Callaway Plant P hae 2 pre-solved table item "Diesel Generator Fails to Run after Start," the inspectors determined this finding to be potentially risk significant. This finding was forwarded to a senior reactor analyst for review. The results of the senior reactor analyst's Phase 3 analysis determined the finding to be of very low safety significance. This finding did not have a crosscutting aspect since it was not a performance deficiency indicative of current licensee performance.
Inspection Report# : 2009007 (pdf)
 
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance:        Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Comply with Radiation Work Permit Requirements The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1.a, which resulted from a failure to comply with radiation work permit instructions. Specifically, on November 2, 2008, during a change out of the chemical and volume control system reactor coolant Filter FBG06, the technicians failed to follow radiation work permit instructions that required notification of the ALARA specialist if the vent port radiation monitor reading was greater than or equal to 1500 millirem per hour to determine if additional briefing requirements were needed. The licensee entered this item into their corrective action program as Callaway Action Request 200811469. As corrective action, the licensee has modified the briefing procedure and modified the radiation work permits to include a requirement to notify radiation protection supervision to evaluate dose rate readings of the vent port and filter housing.
Other corrective actions are being evaluated.
Failure to comply with radiation work permit requirements is a performance deficiency. The finding is greater than minor because it is associated with the cornerstone attribute of exposure control and affected the cornerstone objective, in that, the failure to follow radiation work permit requirements increases the potential for increased dose.
The finding involved workers unplanned, unintended doses or potential of such a dose (resulting from actions or conditions contrary to the radiation work permit). Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, the finding had a crosscutting aspect in the area of human performance, work practices, because the licensee failed to communicate human error prevention techniques during the prejob briefing and ensure that all personnel understood limits stated in the radiation work permit. In addition, personnel proceeded with the filter change out even though radiation levels were significantly higher than anticipated.
Inspection Report# : 2009003 (pdf)
Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous
 
Last modified : March 01, 2010 Callaway 1Q/2010 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Plant Transient Caused by Human Error During Power Range Nuclear Instrument Surveillance The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after maintenance on power range nuclear instrument N41 resulted in an unanticipated plant transient. On October 6, 2009, the licensee performed Procedure ISL-SE-00N41 to calibrate power range nuclear instrument N41. During performance of the test, control rods unexpectedly inserted ten and a half steps at a rate of 72 steps per minute. The negative reactivity that was inserted due to the inward rod motion caused reactor power to drop approximately one percent power and pressurizer pressure to drop from 2235 psig to approximately 2223 psig. Subsequent review by the licensee determined that the cause of the undesired rod motion was the rod bank selector switch being left in auto rather than other than auto as required by the procedure. The licensee initiated Callaway Action Request 200908596 to address the causes of the unanticipated plant transient.
This finding was determined to be greater than minor because it impacted the Initiating Events Cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the technical specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the reactor operator who failed to place the rod bank selector switch into the procedurally required position failed to use human error prevention techniques, such as self- and peer-checking [H.4(a)].
Inspection Report# : 2009005 (pdf)
Mitigating Systems Significance:      Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain and Adequate Flooding Analysis The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after AmerenUE failed to provide adequate design control measures for verifying the adequacy of flooding analysis for the auxiliary feedwater pipe chase room 1206/1207. The revised calculation, performed on December 4, 2001, determined that the 10-inch piping from the condensate storage tank going to the main condenser was the limiting source of potential flooding. However several missing or incorrect assumptions challenged the basis for operability of safety related auxiliary feedwater pump transmitters located in the room 22 inches above the floor level. On December 16, 2009, the licensee reperformed the flooding analysis calculation, M-FL-04, Revision 5, including the main condenser as an additional source of flooding. Although 984 gpm of margin was lost due to inclusion of the condenser as a source, the revised analysis supported an operability determination for the transmitters as operable.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability
 
of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time, and did not increase the likelihood of a seismic, flooding, or severe weather initiating event. This finding was determined to not have a crosscutting aspect as the calculation of record was not reflective of current licensee performance.
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of Failure to Follow Operability Determination Procedure The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for two examples of failure to follow Procedure APA-ZZ-00500, Appendix 1, Operability and Functionality Determinations. The first example occurred on January 14, 2009, following an immediate operability determination made in response to Callaway Action Request 200900231. That Callaway action request documented significant emergency diesel generator heat exchanger tube wall thinning during eddy current testing. The operability determination performed in response to the degraded condition identified in Callaway Action Request 200900231 assumed a linear rate of degradation based on the rate observed from 2006 to 2008 and extrapolated forward to predict when heat exchanger tube plugging limits would be exceeded. Subsequent eddy current testing by the licensee found that the assumed linear degradation rate was nonconservative. The inspectors determined that the licensee failed to provide a reasonable expectation of operability consistent with the requirements of licensee Procedure APA-ZZ-00500, Appendix 1. Specifically, the licensee assumed a nonconservative linear rate of degradation for demonstrating emergency diesel heat exchanger operability despite empirical data that suggested the rate increased as a function of time.
The second example occurred on December 10, 2009, following initiation of Callaway Action Request 200910153 which documented that the steam generator C atmospheric steam dump valve (ABPV0003) would not repeatedly stroke to the same position. The Callaway action request documented that some amount of foreign material within the valve positioner was the cause of the repeatability issue with the valve. The inspectors reviewed Callaway Action Request 200910153 and noted that an immediate operability determination was not made on the identified degraded condition of foreign material within the air supply to the steam generator atmospheric steam dump valves. Since all four steam generator atmospheric steam dump valves share a common instrument air supply, the inspectors determined that the licensee failed to identify what structures, systems, and components were affected by the degraded condition in Callaway Action Request 200910153. Following questioning by the inspectors, the licensee tested the remaining three steam generator atmospheric steam dump valves. During that testing, the licensee found the steam generator B atmospheric steam dump valve would not consistently stroke and that there was a small amount of foreign material within the air operated valve positioner.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations [H.1(b)].
Inspection Report# : 2009005 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Turbine-driven auxiliary feedwater pump inoperable due to inadequately lubricated trip throttle valve
 
The team identified a self-revealing apparent violation of Technical Specification 3.7.5, Auxiliary Feedwater System, due to the failure to adequately lubricate turbine-driven auxiliary feedwater pump trip throttle valve FCHV0312. During May 25, 2009, surveillance testing, the turbine-driven auxiliary feedwater pump did not start as expected due to hardened grease on the valve spindle of FCHV0312. The previous lubrication preventative maintenance had been missed and lack of lubrication increased friction between the sliding nut and spindle preventing FCHV0312 from opening. Following lubrication FCHV0312 and the turbine-driven auxiliary feedwater pump tested satisfactorily. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200904216.
This finding is greater than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as potentially risk significant since the finding represented a loss of system safety function because the turbine-driven auxiliary feedwater pump PAL02 failing eliminates the capability of the plant to cope with a station blackout. The finding required a Phase 2 analysis. When evaluated per Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and the Callaway Plant Phase 2 pre-solved table item Turbine Driven Auxiliary Feedwater Pump Fails to Start, the inspectors determined this finding to be potentially risk significant. The finding was forwarded to a senior reactor analyst for review. The preliminary outcome of the Phase 3 significance determination analysis, Attachment 4, determined the finding was of low to moderate safety significance.
The inspectors determined that this finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to follow the procedural guidance provided when changing the scope of a preventive maintenance task [H.4(b)].
Inspection Report# : 2009009 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Lubrication Procedure for Valve FCHV0312 The team identified a noncited violation of Technical Specification 5.4.1.a, Procedures, for the failure to provide adequate procedural guidance for the lubrication of auxiliary feedwater pump turbine trip throttle valve FCHV0312.
The inspectors found that 2002 corrective actions to improve the lubrication procedure were not fully developed and the procedure lubrication guidance was ambiguous in that it did not specify the amount of lubricant to apply or what valve subcomponents to lubricate. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905032.
This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not have a crosscutting aspect since the 2003 lubrication procedure revision was not reflective of current licensee performance.
Inspection Report# : 2009009 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately evaluate the use of Mobile 28 grease for the turbine-driven auxiliary feedwater pump trip throttle valve.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the
 
failure to adequately evaluate the use of Mobile 28 grease for the turbine-driven auxiliary feedwater pump trip throttle valve. The licensees 1995 evaluation included no documentation for the appropriate relubrication interval of the valve. Additionally, the inspectors identified that the valve exhibited temperatures ranging from 235°F to near 300°F compared to the 215°F valve temperature used in the evaluation. The inspectors questioned if the use of Mobile 28 grease was appropriate since operating experience suggests that Mobile 28 grease has a tendency to thicken and harden at temperatures exceeding 250°F and elevated temperatures increased the lubricants tendency to lose oils and could result in increased stem friction. Following questioning by the inspectors, the licensee initiated Callaway Action Request 200905067 and Request for Resolution 200905651 to determine if Mobile 28 grease was an appropriate lubricant for valve FCHV0312.
This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not to have a crosscutting aspect since the inadequate 1995 lubrication evaluation was not reflective of current licensee performance.
Inspection Report# : 2009009 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to enter conditions adverse to quality associated with the turbine-driven auxiliary feedwater pump trip throttle valve into the corrective action program.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Callaway Procedure APA-ZZ-00500, Corrective Action Program. Specifically, licensee personnel failed to initiate Callaway action requests for adverse conditions of high hand wheel forces, galled subcomponents, and hardened, gritty grease found during the 2007 rebuild of the spare turbine-driven auxiliary feedwater pump trip throttle valve FCHV0312. The licensee has entered this issue into their corrective action program as Callaway Action Request 200905053.
This finding is greater than minor because, if left uncorrected, failure to fully utilize the corrective action program could become a more significant safety concern. The inspectors determined that this finding impacted the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The cause of this finding is related to the problem identification and resolution crosscutting component of the corrective action program because licensee personnel failed to implement a corrective action program with a low threshold for identifying issues
[P.1(a)].
Inspection Report# : 2009009 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure turbine-driven auxiliary feedwater pump is operable prior to entry into Mode 3.
The team identified a noncited violation of Technical Specification Limiting Condition for Operation 3.0.4 for entering Mode 3 with the turbine-driven auxiliary feedwater pump inoperable. Specifically, on November 3, 2008, while in Mode 4 for Refueling Outage 16, an unexpected overspeed trip of the turbine occurred during postmaintenance testing. Callaway operations staff inappropriately concluded that a water slug from the auxiliary
 
steam line was the cause of the turbine overspeed. Following entry into Mode 3, during preparations for turbine-driven auxiliary feedwater pump testing, the licensee found the servo control valve installed during the outage was faulty. When questioned by the inspectors, the licensee determined that the faulty servo control valve discovered in Mode 3 was responsible for the overspeed of the turbine-driven auxiliary feedwater pump that occurred in Mode 4 and that the equipment was inoperable during the mode change that occurred on November 4, 2008. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905313.
This finding is greater than minor because it is associated with the Mitigating Systems Cornerstone attribute of equipment performance and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The inspectors determined that this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to fully evaluate the overspeed of the turbine-driven auxiliary feedwater pump that occurred on November 3, 2008 [P.1(c)].
Inspection Report# : 2009009 (pdf)
Significance:      Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions for Essential Service Water Pump Cable Underground Electrical Vault Seals The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
associated with the licensees failure to take prompt corrective actions to prevent continuous submergence of essential service water pump kerite insulated power cables. The continuously submerged environment for these cables existed because the two vaults containing these cables (MH-01N and MH-01S) had inadequate seals needed to protect the vaults from incoming surface water. Callaway Action Request 200201916 stated that all medium voltage cables of concern were located more than 4 feet above the basemat of the vault and thus were not in a submerged condition. The Callaway action request noted that the seals at the top lid were the source of water intrusion and that the seal design was inadequate. On July 9 and 22, 2009, the resident inspectors, along with Callaway plant engineers, inspected the two essential service water underground vaults. The north vault (train A) was found to have water covering the two safety related upper cable trays. Contrary to the Callaway Action Request 200201916 evaluation, the cable trays were about 2.5 feet and 3.5 feet from the basemat of the vault floor. During these 2009 inspections it was noted that the same lid design deficiency identified in Callaway Action Request 200201916 still existed. This led to the discovery that the Callaway Action Request actions from 2002 had not been completely performed. The only significant corrective action had been to increase the inspection frequency to once every three years.
The licensee has subsequently taken measures to improve the seals and written Callaway Action Request/Request for Resolution 200905838 to further evaluate this issue. This finding is more than minor because it affected the Mitigating Systems Cornerstone attribute of design control for ensuring the availability, reliability, and capability of safety systems. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded seals were a design or qualification deficiency confirmed not to result in loss of operability. The inspectors determined that the finding has no crosscutting aspect as the performance deficiencies were not reflective of current performance. The licensee entered this item into their corrective action program as Callaway Action Request 200908855.
Inspection Report# : 2009004 (pdf)
Significance: SL-IV Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correctly Identify Safety System Functional Failures in a Licensee Event Report The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for a failure to report two examples of safety system functional failures in licensee event reports within 60
 
days after discovery of events requiring a report. The two examples were:
* March 26, 2008, discovery that operation of containment air coolers in fast speed, during a period of higher than normal containment pressure, could open the air coolers fast speed thermal overload device rendering all the coolers incapable of automatically restarting in slow speed
* May 21, 2008, discovery of a 6.6 cubic foot void of air in the common suction piping capable of affecting the function of both of the safety injection system pumps For each example, the inspectors reviewed the licensees reportability evaluation and associated past operability reviews and determined each event was reportable per 10 CFR 50.73(a)(2)(v) since each example resulted in a condition which affected both trains of a system described in the Final Safety Analysis Report that was needed to mitigate the consequences of an accident. Alternate safety systems accident mitigation is not permitted as a reason to not report the discovery of the conditions. The licensee also failed to report these failures to the NRC performance indicator database because of the failure to include the safety system functional failure in each respective licensee event report.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. The licensee planned to update the associated license event reports as described in Callaway Action Request 200904980. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure, through adequate training, that its staff understood the guidance documents pertaining to the 10 CFR 50.73 rule [H.2.(b)].
Inspection Report# : 2009004 (pdf)
Significance: SL-IV Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit Complete and Accurate Risk Information for a Requested License Amendment The inspectors identified a noncited violation of 10 CFR 50.9, "Completeness and Accuracy of Information," when AmerenUE failed to submit complete and accurate quantification of risk contributors associated with a license amendment supporting a modification to replace the underground portion of the essential service water system Train B piping with high density polyethelene pipe. The inspectors questioned the risk impact of a possible control room fire which led to the discovery that the licensee had not followed their process for screening out fire areas. The licensee entered this item into their corrective action program as Callaway Action Request 200902810 and also submitted an update to License Amendment 191 to correctly account for the control room fire risk.
This finding affects the Mitigating Systems cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect because the licensees failure to thoroughly review and submit the risk for control room fires was not part of a corrective action process, but instead an oversight by the licensing review process.
Inspection Report# : 2009003 (pdf)
Significance:      Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Controls of Crane Work Above the Protected Train of Essential Service Water The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) associated with the licensees failure to adequately assess and manage risk associated with crane work over the essential service water system Train A. On March 31, 2009, the licensee performed work in the vicinity of the protected essential service water system train which included movement of 1800 pound sand bags over the protected train piping. After questioning by the resident
 
inspectors, the licensee determined that the lifts were not conducted in accordance with station procedures since the requirements of a required engineering judgment memo were not translated into work documents. The licensee entered this item into their corrective action program as Callaway Action Request 200902726.
The finding affected the Mitigating Systems cornerstone and was determined to be more than minor because the licensee failed to implement the prescribed significant compensatory measures associated with crane work in the vicinity of safe shutdown equipment. This finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to include appropriate risk insights in planned work activities.
Inspection Report# : 2009003 (pdf)
Significance:      Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate, At Power, Risk Assessment for Maintenance Activities on One Train of Essential Service Water and Emergency Diesel Generator The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) associated with the licensees failure to perform an adequate risk assessment for planned maintenance on the emergency diesel generator Train A and essential service water pump Train A. On April 28, 2009, Callaway Plant operators removed the emergency diesel generator Train A and essential service water pump Train A from service. The inspectors' review of the plant risk profile for the in-progress maintenance activity uncovered that this risk had not been accounted for by the plant safety monitor tool.
The licensee entered this item into their corrective action program as Callaway Action Request 200903480 The finding is more than minor because the risk, when correctly assessed, put the plant into a higher risk category for large early release frequency. Also the licensee risk assessment failed to consider risk significant systems, structures, and components and support systems that were unavailable during the maintenance. This finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to appropriately plan work activities consistent with nuclear safety by incorporating risk insights.
Inspection Report# : 2009003 (pdf)
Significance:      Apr 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Suitable Replacement Parts Essential for Emergency Diesel Generator Train B The inspectors identifed a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" after the licensee failed to adequately select suitable replacement parts essential to the operation of emergency diesel generator Train B. On December 24, 2008, during performance of Procedure OSP-NE-0001B, "Standby Diesel Generator B Periodic Tests, " Callaway operations personnel identified that the emergency diesel generator Train B had an approximately 0.82 gallon per minute jacket water leak resulting in operators declaring the equipment inoperable.
Upon removal, the gasket was found to be soft and extruding from the flange edge. The licensee originally concluded the gasket failed due to vibrations associated with engine shutdown but altered that conclusion after discussions witht he resident inspectors and additional investigation. The licensee ultimately determined that the cause of the failure was due to incorrect gasket material being used during Job W200773 performed on October 16, 1999. The gasket was 1/8" thick which resulted ina lack of compression. Since the gaskets are composed of an aramid fiberous material, the lack of compression allowed the gasket to absorb water and soften. The leak identified on December 24, 2008, developed once the gasket softened sufficiently to extrude from the flange edge. This issue has been entered into the licensee's corrective action program as Callaway Action Request 200812985.
This finding was greater than minor because it was associated with the mitigating systems cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings, " this finding was determined to represent an actual loss of safety function of a single train for greater thatn its Technical Specification allowed outage time. When evaluated per Manual Chapter 0609 Appendiz A, "Determining the Significance of Reactor Inspection Finding for At-Power
 
Situations," and the Callaway Plant P hae 2 pre-solved table item "Diesel Generator Fails to Run after Start," the inspectors determined this finding to be potentially risk significant. This finding was forwarded to a senior reactor analyst for review. The results of the senior reactor analyst's Phase 3 analysis determined the finding to be of very low safety significance. This finding did not have a crosscutting aspect since it was not a performance deficiency indicative of current licensee performance.
Inspection Report# : 2009007 (pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance:        Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Comply with Radiation Work Permit Requirements The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1.a, which resulted from a failure to comply with radiation work permit instructions. Specifically, on November 2, 2008, during a change out of the chemical and volume control system reactor coolant Filter FBG06, the technicians failed to follow radiation work permit instructions that required notification of the ALARA specialist if the vent port radiation monitor reading was greater than or equal to 1500 millirem per hour to determine if additional briefing requirements were needed. The licensee entered this item into their corrective action program as Callaway Action Request 200811469. As corrective action, the licensee has modified the briefing procedure and modified the radiation work permits to include a requirement to notify radiation protection supervision to evaluate dose rate readings of the vent port and filter housing.
Other corrective actions are being evaluated.
Failure to comply with radiation work permit requirements is a performance deficiency. The finding is greater than minor because it is associated with the cornerstone attribute of exposure control and affected the cornerstone objective, in that, the failure to follow radiation work permit requirements increases the potential for increased dose.
The finding involved workers unplanned, unintended doses or potential of such a dose (resulting from actions or conditions contrary to the radiation work permit). Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, the finding had a crosscutting aspect in the area of human performance, work practices, because the licensee failed to communicate human error prevention techniques during the prejob briefing and ensure that all personnel understood limits stated in the radiation work permit. In addition, personnel proceeded with the filter change out even though radiation levels were significantly higher than anticipated.
Inspection Report# : 2009003 (pdf)
Public Radiation Safety Physical Protection
 
Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : May 26, 2010
 
Callaway 2Q/2010 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2010 Identified By: NRC Item Type: FIN Finding Failure to Ensure Completion of Corrective Actions for Degraded Chemical and Volume Control System Valves The inspectors identified a finding associated with AmerenUEs failure to take prompt corrective actions for leaking boundary valves in the chemical and volume control system. On April 13, 2010, an attempt to place the train A chemical and volume control system mixed bed in service resulted in leakage past a documented leaking drain valve.
The lingering equipment problems resulted in an unplanned 25 gallon per minute loss rate of volume control tank inventory and an emergency action level declaration for excessive reactor coolant system leakage. Later, the declaration was retracted. The licensee placed this issue into the corrective action program as Callaway Action Request 201003146.
This finding is more than minor because it was associated with the reactor safety Initiating Events Cornerstone attribute of configuration control and affected the objective to limit the likelihood of events that upset plant stability.
Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding is of very low significance because the condition did not result in the reactor coolant system technical specification leakage limit being exceeded, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or internal/external flood. This finding, which involved inadequate scheduling of corrective action related jobs, has a crosscutting aspect in the area of human performance associated with the work control component because AmerenUE did not appropriately coordinate work activities to address the impact of the work on different job activities.
Inspection Report# : 2010003 (pdf)
Significance:        Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Plant Transient Caused by Human Error During Power Range Nuclear Instrument Surveillance The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after maintenance on power range nuclear instrument N41 resulted in an unanticipated plant transient. On October 6, 2009, the licensee performed Procedure ISL-SE-00N41 to calibrate power range nuclear instrument N41. During performance of the test, control rods unexpectedly inserted ten and a half steps at a rate of 72 steps per minute. The negative reactivity that was inserted due to the inward rod motion caused reactor power to drop approximately one percent power and pressurizer pressure to drop from 2235 psig to approximately 2223 psig. Subsequent review by the licensee determined that the cause of the undesired rod motion was the rod bank selector switch being left in auto rather than other than auto as required by the procedure. The licensee initiated Callaway Action Request 200908596 to address the causes of the unanticipated plant transient.
This finding was determined to be greater than minor because it impacted the Initiating Events Cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the technical specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the reactor operator who failed to place
 
the rod bank selector switch into the procedurally required position failed to use human error prevention techniques, such as self- and peer-checking [H.4(a)].
Inspection Report# : 2009005 (pdf)
Mitigating Systems Significance:        Jun 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Surveillance Procedure to Verify and Maintain Emergency Core Cooling System Operable The inspectors identified a noncited violation of Technical Specification 3.5.2, Emergency Core Cooling Systems.
Specifically Technical Specifications Surveillance Requirement 3.5.2.3, Verify the ECCS piping is full of water, was not being met by licensee Procedure OSP SA 00003, Emergency Core Cooling System Flow Path Verification and Venting. On April 22, 2010, the inspectors discovered that the train B residual heat removal system discharge line EJ 024 ECB 10 did not have an accessible high point vent. The line was required by Callaway procedures to be either monitored by venting or tested using an ultrasonic method as described in the procedures acceptance criteria.
Callaway had identified the need to install a vent valve in line EJ 024 ECB-10 per modification MP 08 0016 prior to Refueling Outage 17. The licensee originally scheduled the vent valve installation during Refueling Outage 17, but had inappropriately deferred the maintenance to the next outage in fall 2011. As immediate corrective action, the licensee installed the vent valves in Refueling Outage 17 and placed this issue into the corrective action program as Callaway Action Request 201004078.
This finding is more than minor because it affected the Mitigating Systems Cornerstone procedure quality attribute and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding is of very low significance because it was only a design or qualification deficiency confirmed not to result in loss of operability. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions in decision making and did not adopt a requirement to demonstrate that either venting or ultrasonic testing was needed to verify line EJ 024 ECB 10 was full of water.
Inspection Report# : 2010003 (pdf)
Significance:        Jun 23, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Fabricate Replacement Gasket for Emergency Diesel Generator TrainA The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to adequately select suitable replacement gaskets essential to the operation of emergency diesel generator train A. On March 30, 2010, during performance of Procedure OSP-NE-00024A, Standby Diesel Generator A 24-Hour Run and Hot Restart Test, the emergency diesel generator train A unexpectedly lost speed and tripped after 16.7 hours of operation. Posttrip indications revealed that the diesel generator tripped from a stripped splined shaft in the governor drive housing. The failure of the splined shaft was caused by an improperly cut gasket which did not have the required oil port hole to allow proper lubrication of the drive assembly. The licensee replaced the damaged shaft and placed this issue in their corrective action program as Callaway Action Request 201002675.
This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The resident inspectors performed the initial significance determination for the diesel gasket finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved the loss of one train of safety related equipment for greater than its
 
technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination using the pre-solved worksheet from the Risk Informed Inspection Notebook for Callaway Nuclear Generating Station, Revision 2.01a. The analyst assumed an exposure period of one year. The finding was potentially Yellow, which warranted further review. The senior reactor analyst subsequently performed a bounding Phase 3 significance determination and found the finding to be of very low safety significance (Green). The dominant cutsets included a loss of offsite power initiating event, failure to recover offsite power in 4 hours, failure of the train B emergency diesel generator, and a reactor coolant pump seal failure. Equipment that mitigated the significance included the operable emergency diesel generator and the turbine-driven auxiliary feedwater pump. This finding did not have a crosscutting aspect since it was not a performance deficiency reflective of current licensee performance.
Inspection Report# : 2010003 (pdf)
Significance:      Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Operability Determination Procedure The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to follow Procedure APA-ZZ-00500, Appendix 1, Operability and Functionality Determinations. The inspectors determined that the licensee failed to provide a reasonable expectation of operability for the degraded condition. Specifically, the licensee failed to account for both auxiliary feedwater as an essential service water system load and fouling resistance in the component cooling water system heat exchanger. Long term corrective actions planned include a modification of the component cooling water heat exchangers divider plate during the upcoming April 2010 refueling outage. The licensee placed this issue in their corrective action program as Callaway Action Request 201001152.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations [H.1(b)].
Inspection Report# : 2010002 (pdf)
Significance:      Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Suitable Replacement Parts Essential for the Operation of the Component Cooling Water System The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to adequately select suitable replacement gaskets essential to the operation of the component cooling water system heat exchangers. On October 19, 2008, Callaway engineering personnel identified that the component cooling water heat exchangers, due to corrosion and inadequate gasket sealing, had a small gap between the divider plate and channel head such that it allowed essential service water flow to bypass the heat exchanger which resulted in a reduced heat transfer capability. Corrective actions to address the identified gap in the component cooling water heat exchanger were scheduled to be implemented during the licensees next refueling outage. The licensee entered the issue in the corrective action program as Callaway Action Request 201001900.
This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not
 
affect seismic, flooding, or severe weather initiating events. This finding was determined not to have a crosscutting aspect since it is a performance deficiency not reflective of current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:        Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to M aintain an Adequare Ultimate Heat Sink Thermal Performance Analysis The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after AmerenUE failed to provide adequate design control measures for verifying the adequacy of the ultimate heat sink thermal performance analysis evaluating the impact of heat rejected during a large break loss of coolant accident. The thermal performance analysis, most recently revised in 2007, did not account for a potential single active failure of each trains motor-operated valve designed to redirect the essential service water return flow up and over the tower fill material. With further analysis the licensee determined that a compensatory measure implementing a more restrictive initial operating range based on pond volume and initial temperature would ensure that the ultimate heat sink pond will not exceed its maximum temperature of 92.3 degrees Fahrenheit during a design basis accident. Corrective actions were being developed using Callaway Action Request 201001813.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A resident inspector performed the initial significance determination for the inoperable essential service water system, under certain conditions, using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved the potential inoperability of both trains of essential service water for greater than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and found that the finding was potentially greater than green. The senior reactor analyst then performed a bounding Phase 3 significance determination and found the finding to be of very low safety significance (Green).
The dominant core damage sequences included a medium break loss of coolant accident concurrent with the failure of essential service water system cooling tower bypass valves. The finding was mitigated because the motor operated valves remained functional throughout the year, which minimized the frequencies for the scenarios of interest. This finding was determined to not have a crosscutting aspect as the calculation of record was not reflective of current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:        Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain and Adequate Flooding Analysis The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after AmerenUE failed to provide adequate design control measures for verifying the adequacy of flooding analysis for the auxiliary feedwater pipe chase room 1206/1207. The revised calculation, performed on December 4, 2001, determined that the 10-inch piping from the condensate storage tank going to the main condenser was the limiting source of potential flooding. However several missing or incorrect assumptions challenged the basis for operability of safety related auxiliary feedwater pump transmitters located in the room 22 inches above the floor level. On December 16, 2009, the licensee reperformed the flooding analysis calculation, M-FL-04, Revision 5, including the main condenser as an additional source of flooding. Although 984 gpm of margin was lost due to inclusion of the condenser as a source, the revised analysis supported an operability determination for the transmitters as operable.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time,
 
and did not increase the likelihood of a seismic, flooding, or severe weather initiating event. This finding was determined to not have a crosscutting aspect as the calculation of record was not reflective of current licensee performance.
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of Failure to Follow Operability Determination Procedure The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for two examples of failure to follow Procedure APA-ZZ-00500, Appendix 1, Operability and Functionality Determinations. The first example occurred on January 14, 2009, following an immediate operability determination made in response to Callaway Action Request 200900231. That Callaway action request documented significant emergency diesel generator heat exchanger tube wall thinning during eddy current testing. The operability determination performed in response to the degraded condition identified in Callaway Action Request 200900231 assumed a linear rate of degradation based on the rate observed from 2006 to 2008 and extrapolated forward to predict when heat exchanger tube plugging limits would be exceeded. Subsequent eddy current testing by the licensee found that the assumed linear degradation rate was nonconservative. The inspectors determined that the licensee failed to provide a reasonable expectation of operability consistent with the requirements of licensee Procedure APA-ZZ-00500, Appendix 1. Specifically, the licensee assumed a nonconservative linear rate of degradation for demonstrating emergency diesel heat exchanger operability despite empirical data that suggested the rate increased as a function of time.
The second example occurred on December 10, 2009, following initiation of Callaway Action Request 200910153 which documented that the steam generator C atmospheric steam dump valve (ABPV0003) would not repeatedly stroke to the same position. The Callaway action request documented that some amount of foreign material within the valve positioner was the cause of the repeatability issue with the valve. The inspectors reviewed Callaway Action Request 200910153 and noted that an immediate operability determination was not made on the identified degraded condition of foreign material within the air supply to the steam generator atmospheric steam dump valves. Since all four steam generator atmospheric steam dump valves share a common instrument air supply, the inspectors determined that the licensee failed to identify what structures, systems, and components were affected by the degraded condition in Callaway Action Request 200910153. Following questioning by the inspectors, the licensee tested the remaining three steam generator atmospheric steam dump valves. During that testing, the licensee found the steam generator B atmospheric steam dump valve would not consistently stroke and that there was a small amount of foreign material within the air operated valve positioner.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations [H.1(b)].
Inspection Report# : 2009005 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Turbine-driven auxiliary feedwater pump inoperable due to inadequately lubricated trip throttle valve The team identified a self-revealing apparent violation of Technical Specification 3.7.5, Auxiliary Feedwater System, due to the failure to adequately lubricate turbine-driven auxiliary feedwater pump trip throttle valve FCHV0312. During May 25, 2009, surveillance testing, the turbine-driven auxiliary feedwater pump did not start as expected due to hardened grease on the valve spindle of FCHV0312. The previous lubrication preventative
 
maintenance had been missed and lack of lubrication increased friction between the sliding nut and spindle preventing FCHV0312 from opening. Following lubrication FCHV0312 and the turbine-driven auxiliary feedwater pump tested satisfactorily. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200904216.
This finding is greater than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as potentially risk significant since the finding represented a loss of system safety function because the turbine-driven auxiliary feedwater pump PAL02 failing eliminates the capability of the plant to cope with a station blackout. The finding required a Phase 2 analysis. When evaluated per Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, and the Callaway Plant Phase 2 pre-solved table item Turbine Driven Auxiliary Feedwater Pump Fails to Start, the inspectors determined this finding to be potentially risk significant. The finding was forwarded to a senior reactor analyst for review. The preliminary outcome of the Phase 3 significance determination analysis, Attachment 4, determined the finding was of low to moderate safety significance.
The inspectors determined that this finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to follow the procedural guidance provided when changing the scope of a preventive maintenance task [H.4(b)].
Inspection Report# : 2009009 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Lubrication Procedure for Valve FCHV0312 The team identified a noncited violation of Technical Specification 5.4.1.a, Procedures, for the failure to provide adequate procedural guidance for the lubrication of auxiliary feedwater pump turbine trip throttle valve FCHV0312.
The inspectors found that 2002 corrective actions to improve the lubrication procedure were not fully developed and the procedure lubrication guidance was ambiguous in that it did not specify the amount of lubricant to apply or what valve subcomponents to lubricate. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905032.
This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not have a crosscutting aspect since the 2003 lubrication procedure revision was not reflective of current licensee performance.
Inspection Report# : 2009009 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately evaluate the use of Mobile 28 grease for the turbine-driven auxiliary feedwater pump trip throttle valve.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to adequately evaluate the use of Mobile 28 grease for the turbine-driven auxiliary feedwater pump trip throttle valve. The licensees 1995 evaluation included no documentation for the appropriate relubrication interval of the valve. Additionally, the inspectors identified that the valve exhibited temperatures ranging from 235°F to near 300°F compared to the 215°F valve temperature used in the evaluation. The inspectors questioned if the use of Mobile 28
 
grease was appropriate since operating experience suggests that Mobile 28 grease has a tendency to thicken and harden at temperatures exceeding 250°F and elevated temperatures increased the lubricants tendency to lose oils and could result in increased stem friction. Following questioning by the inspectors, the licensee initiated Callaway Action Request 200905067 and Request for Resolution 200905651 to determine if Mobile 28 grease was an appropriate lubricant for valve FCHV0312.
This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding did not to have a crosscutting aspect since the inadequate 1995 lubrication evaluation was not reflective of current licensee performance.
Inspection Report# : 2009009 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to enter conditions adverse to quality associated with the turbine-driven auxiliary feedwater pump trip throttle valve into the corrective action program.
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Callaway Procedure APA-ZZ-00500, Corrective Action Program. Specifically, licensee personnel failed to initiate Callaway action requests for adverse conditions of high hand wheel forces, galled subcomponents, and hardened, gritty grease found during the 2007 rebuild of the spare turbine-driven auxiliary feedwater pump trip throttle valve FCHV0312. The licensee has entered this issue into their corrective action program as Callaway Action Request 200905053.
This finding is greater than minor because, if left uncorrected, failure to fully utilize the corrective action program could become a more significant safety concern. The inspectors determined that this finding impacted the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The cause of this finding is related to the problem identification and resolution crosscutting component of the corrective action program because licensee personnel failed to implement a corrective action program with a low threshold for identifying issues
[P.1(a)].
Inspection Report# : 2009009 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to ensure turbine-driven auxiliary feedwater pump is operable prior to entry into Mode 3.
The team identified a noncited violation of Technical Specification Limiting Condition for Operation 3.0.4 for entering Mode 3 with the turbine-driven auxiliary feedwater pump inoperable. Specifically, on November 3, 2008, while in Mode 4 for Refueling Outage 16, an unexpected overspeed trip of the turbine occurred during postmaintenance testing. Callaway operations staff inappropriately concluded that a water slug from the auxiliary steam line was the cause of the turbine overspeed. Following entry into Mode 3, during preparations for turbine-driven auxiliary feedwater pump testing, the licensee found the servo control valve installed during the outage was faulty. When questioned by the inspectors, the licensee determined that the faulty servo control valve discovered in Mode 3 was responsible for the overspeed of the turbine-driven auxiliary feedwater pump that occurred in Mode 4 and
 
that the equipment was inoperable during the mode change that occurred on November 4, 2008. The licensee entered this deficiency in their corrective action program as Callaway Action Request 200905313.
This finding is greater than minor because it is associated with the Mitigating Systems Cornerstone attribute of equipment performance and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The inspectors determined that this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to fully evaluate the overspeed of the turbine-driven auxiliary feedwater pump that occurred on November 3, 2008 [P.1(c)].
Inspection Report# : 2009009 (pdf)
Significance:      Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions for Essential Service Water Pump Cable Underground Electrical Vault Seals The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
associated with the licensees failure to take prompt corrective actions to prevent continuous submergence of essential service water pump kerite insulated power cables. The continuously submerged environment for these cables existed because the two vaults containing these cables (MH-01N and MH-01S) had inadequate seals needed to protect the vaults from incoming surface water. Callaway Action Request 200201916 stated that all medium voltage cables of concern were located more than 4 feet above the basemat of the vault and thus were not in a submerged condition. The Callaway action request noted that the seals at the top lid were the source of water intrusion and that the seal design was inadequate. On July 9 and 22, 2009, the resident inspectors, along with Callaway plant engineers, inspected the two essential service water underground vaults. The north vault (train A) was found to have water covering the two safety related upper cable trays. Contrary to the Callaway Action Request 200201916 evaluation, the cable trays were about 2.5 feet and 3.5 feet from the basemat of the vault floor. During these 2009 inspections it was noted that the same lid design deficiency identified in Callaway Action Request 200201916 still existed. This led to the discovery that the Callaway Action Request actions from 2002 had not been completely performed. The only significant corrective action had been to increase the inspection frequency to once every three years.
The licensee has subsequently taken measures to improve the seals and written Callaway Action Request/Request for Resolution 200905838 to further evaluate this issue. This finding is more than minor because it affected the Mitigating Systems Cornerstone attribute of design control for ensuring the availability, reliability, and capability of safety systems. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded seals were a design or qualification deficiency confirmed not to result in loss of operability. The inspectors determined that the finding has no crosscutting aspect as the performance deficiencies were not reflective of current performance. The licensee entered this item into their corrective action program as Callaway Action Request 200908855.
Inspection Report# : 2009004 (pdf)
Significance: SL-IV Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correctly Identify Safety System Functional Failures in a Licensee Event Report The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for a failure to report two examples of safety system functional failures in licensee event reports within 60 days after discovery of events requiring a report. The two examples were:
* March 26, 2008, discovery that operation of containment air coolers in fast speed, during a period of higher than normal containment pressure, could open the air coolers fast speed thermal overload device rendering all the coolers
 
incapable of automatically restarting in slow speed
* May 21, 2008, discovery of a 6.6 cubic foot void of air in the common suction piping capable of affecting the function of both of the safety injection system pumps For each example, the inspectors reviewed the licensees reportability evaluation and associated past operability reviews and determined each event was reportable per 10 CFR 50.73(a)(2)(v) since each example resulted in a condition which affected both trains of a system described in the Final Safety Analysis Report that was needed to mitigate the consequences of an accident. Alternate safety systems accident mitigation is not permitted as a reason to not report the discovery of the conditions. The licensee also failed to report these failures to the NRC performance indicator database because of the failure to include the safety system functional failure in each respective licensee event report.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. The licensee planned to update the associated license event reports as described in Callaway Action Request 200904980. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure, through adequate training, that its staff understood the guidance documents pertaining to the 10 CFR 50.73 rule [H.2.(b)].
Inspection Report# : 2009004 (pdf)
Barrier Integrity Significance:      Jun 23, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Maintain Two Operable Source Range Channels During Core Alterations The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, when the licensees inadequate procedure and failure to control work activities during a reload of the reactor vessel fuel assemblies resulted in deenergization of all available source range nuclear instrument channels. On May 6, 2010, while in Mode 6 - Refueling, licensee testing of nuclear instrument power range channel N44 and maintenance on 120 Vac instrument bus NN03 affecting power range channel N43 made up the logic for permissive P 10. The permissive sent a protective logic signal to deenergize both available source range nuclear instruments. The control room immediately directed the fuel handling crew to stop fuel movement until the source range channels could be restored. A fuel assembly was in the upender ready for transfer to the reactor vessel core location at the time. The licensee placed this issue into the corrective action program as Callaway Action Request 201004301.
This finding is more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or releases. Using Manual Chapter 0609 Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 - Operational Checklists for Both PWRs and BWRs, this finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory, did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and did not degrade the licensees ability to recover decay heat removal once lost. This finding had a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to coordinate work activities by incorporating actions to address the impact of the work on different job activities and communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2010003 (pdf)
 
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : September 02, 2010
 
Callaway 3Q/2010 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2010 Identified By: NRC Item Type: FIN Finding Failure to Ensure Completion of Corrective Actions for Degraded Chemical and Volume Control System Valves The inspectors identified a finding associated with AmerenUEs failure to take prompt corrective actions for leaking boundary valves in the chemical and volume control system. On April 13, 2010, an attempt to place the train A chemical and volume control system mixed bed in service resulted in leakage past a documented leaking drain valve.
The lingering equipment problems resulted in an unplanned 25 gallon per minute loss rate of volume control tank inventory and an emergency action level declaration for excessive reactor coolant system leakage. Later, the declaration was retracted. The licensee placed this issue into the corrective action program as Callaway Action Request 201003146.
This finding is more than minor because it was associated with the reactor safety Initiating Events Cornerstone attribute of configuration control and affected the objective to limit the likelihood of events that upset plant stability.
Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding is of very low significance because the condition did not result in the reactor coolant system technical specification leakage limit being exceeded, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or internal/external flood. This finding, which involved inadequate scheduling of corrective action related jobs, has a crosscutting aspect in the area of human performance associated with the work control component because AmerenUE did not appropriately coordinate work activities to address the impact of the work on different job activities [H.3(b)].
Inspection Report# : 2010003 (pdf)
Significance:        Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Plant Transient Caused by Human Error During Power Range Nuclear Instrument Surveillance The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after maintenance on power range nuclear instrument N41 resulted in an unanticipated plant transient. On October 6, 2009, the licensee performed Procedure ISL-SE-00N41 to calibrate power range nuclear instrument N41. During performance of the test, control rods unexpectedly inserted ten and a half steps at a rate of 72 steps per minute. The negative reactivity that was inserted due to the inward rod motion caused reactor power to drop approximately one percent power and pressurizer pressure to drop from 2235 psig to approximately 2223 psig. Subsequent review by the licensee determined that the cause of the undesired rod motion was the rod bank selector switch being left in auto rather than other than auto as required by the procedure. The licensee initiated Callaway Action Request 200908596 to address the causes of the unanticipated plant transient.
This finding was determined to be greater than minor because it impacted the Initiating Events Cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the technical specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This finding has a crosscutting aspect in the area of human performance associated with the work practices component because the reactor operator who failed to place
 
the rod bank selector switch into the procedurally required position failed to use human error prevention techniques, such as self- and peer-checking [H.4(a)].
Inspection Report# : 2009005 (pdf)
Mitigating Systems Significance: SL-IV Sep 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Accurately Report a Condition that Could Have Prevented Fulfillment of a Safety Function The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), "Licensee Event Report System," for failure to report simultaneous inoperability of two steam generator atmospheric steam dump valves as a condition that could have prevented fulfillment of a safety function. On February 8, 2010, AmerenUE submitted Licensee Event Report 05000483/2009-005-00 to document that steam generator atmospheric steam dump valve ABPV0002 was out of service longer than allowed by Technical Specification 3.7.4, "Atmospheric Steam Dump Valves (ASDs)." The licensee event report also documented a period where valve ABPV0002 inoperability overlapped the inoperability of steam generator atmospheric steam dump valve ABPV0003. Callaway Final Safety Analysis Report Section 15.6.3.2.2.p. stated that all three intact steam generator atmospheric steam dump valves are credited in the cool down for a steam generator tube rupture. The inspectors determined that the licensee failed to adequately evaluate the reportability of having simultaneous inoperability of two steam generator atmospheric steam dump valves as a safety system functional failure. This issue was entered into the licensees corrective action program as Callaway Action Request 201006086 and on September 29, 2010, the licensee submitted Licensee Event Report 05000483/2009-005-001 to correct the reporting error.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was strictly associated with a traditional enforcement violation.
Inspection Report# : 2010004 (pdf)
Significance:      Jun 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Surveillance Procedure to Verify and Maintain Emergency Core Cooling System Operable The inspectors identified a noncited violation of Technical Specification 3.5.2, Emergency Core Cooling Systems.
Specifically Technical Specifications Surveillance Requirement 3.5.2.3, Verify the ECCS piping is full of water, was not being met by licensee Procedure OSP-SA-00003, Emergency Core Cooling System Flow Path Verification and Venting. On April 22, 2010, the inspectors discovered that the train B residual heat removal system discharge line EJ-024-ECB-10' did not have an accessible high point vent. The line was required by Callaway procedures to be either monitored by venting or tested using an ultrasonic method as described in the procedures acceptance criteria.
Callaway had identified the need to install a vent valve in line EJ-024-ECB-10' per modification MP-08-0016 prior to Refueling Outage 17. The licensee originally scheduled the vent valve installation during Refueling Outage 17, but had inappropriately deferred the maintenance to the next outage in fall 2011. As immediate corrective action, the licensee installed the vent valves in Refueling Outage 17 and placed this issue into the corrective action program as Callaway Action Request 201004078.
This finding is more than minor because it affected the Mitigating Systems Cornerstone procedure quality attribute and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding is of very low significance because it was only a design or qualification deficiency confirmed not to result in loss of operability. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the
 
licensee failed to use conservative assumptions in decision making and did not adopt a requirement to demonstrate that either venting or ultrasonic testing was needed to verify line EJ-024-ECB-10 was full of water [H.1(b)].
Inspection Report# : 2010003 (pdf)
Significance:        Jun 23, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Fabricate Replacement Gasket for Emergency Diesel Generator TrainA The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to adequately select suitable replacement gaskets essential to the operation of emergency diesel generator train A. On March 30, 2010, during performance of Procedure OSP-NE-00024A, Standby Diesel Generator A 24-Hour Run and Hot Restart Test, the emergency diesel generator train A unexpectedly lost speed and tripped after 16.7 hours of operation. Posttrip indications revealed that the diesel generator tripped from a stripped splined shaft in the governor drive housing. The failure of the splined shaft was caused by an improperly cut gasket which did not have the required oil port hole to allow proper lubrication of the drive assembly. The licensee replaced the damaged shaft and placed this issue in their corrective action program as Callaway Action Request 201002675.
This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The resident inspectors performed the initial significance determination for the diesel gasket finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved the loss of one train of safety related equipment for greater than its technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination using the pre-solved worksheet from the Risk Informed Inspection Notebook for Callaway Nuclear Generating Station, Revision 2.01a. The analyst assumed an exposure period of one year. The finding was potentially Yellow, which warranted further review. The senior reactor analyst subsequently performed a bounding Phase 3 significance determination and found the finding to be of very low safety significance (Green). The dominant cutsets included a loss of offsite power initiating event, failure to recover offsite power in 4 hours, failure of the train B emergency diesel generator, and a reactor coolant pump seal failure. Equipment that mitigated the significance included the operable emergency diesel generator and the turbine-driven auxiliary feedwater pump. This finding did not have a crosscutting aspect since it was not a performance deficiency reflective of current licensee performance.
Inspection Report# : 2010003 (pdf)
Significance:        Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Operability Determination Procedure The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to follow Procedure APA-ZZ-00500, Appendix 1, Operability and Functionality Determinations. The inspectors determined that the licensee failed to provide a reasonable expectation of operability for the degraded condition. Specifically, the licensee failed to account for both auxiliary feedwater as an essential service water system load and fouling resistance in the component cooling water system heat exchanger. Long term corrective actions planned include a modification of the component cooling water heat exchangers divider plate during the upcoming April 2010 refueling outage. The licensee placed this issue in their corrective action program as Callaway Action Request 201001152.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification
 
allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations [H.1(b)].
Inspection Report# : 2010002 (pdf)
Significance:        Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Suitable Replacement Parts Essential for the Operation of the Component Cooling Water System The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to adequately select suitable replacement gaskets essential to the operation of the component cooling water system heat exchangers. On October 19, 2008, Callaway engineering personnel identified that the component cooling water heat exchangers, due to corrosion and inadequate gasket sealing, had a small gap between the divider plate and channel head such that it allowed essential service water flow to bypass the heat exchanger which resulted in a reduced heat transfer capability. Corrective actions to address the identified gap in the component cooling water heat exchanger were scheduled to be implemented during the licensees next refueling outage. The licensee entered the issue in the corrective action program as Callaway Action Request 201001900.
This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding was determined not to have a crosscutting aspect since it is a performance deficiency not reflective of current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:        Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to M aintain an Adequare Ultimate Heat Sink Thermal Performance Analysis The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after AmerenUE failed to provide adequate design control measures for verifying the adequacy of the ultimate heat sink thermal performance analysis evaluating the impact of heat rejected during a large break loss of coolant accident. The thermal performance analysis, most recently revised in 2007, did not account for a potential single active failure of each trains motor-operated valve designed to redirect the essential service water return flow up and over the tower fill material. With further analysis the licensee determined that a compensatory measure implementing a more restrictive initial operating range based on pond volume and initial temperature would ensure that the ultimate heat sink pond will not exceed its maximum temperature of 92.3 degrees Fahrenheit during a design basis accident. Corrective actions were being developed using Callaway Action Request 201001813.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A resident inspector performed the initial significance determination for the inoperable essential service water system, under certain conditions, using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved the potential inoperability of both trains of essential service water for greater than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and found that the finding was potentially greater than green. The senior reactor analyst then performed a bounding Phase 3 significance determination and found the finding to be of very low safety significance (Green).
The dominant core damage sequences included a medium break loss of coolant accident concurrent with the failure of essential service water system cooling tower bypass valves. The finding was mitigated because the motor operated
 
valves remained functional throughout the year, which minimized the frequencies for the scenarios of interest. This finding was determined to not have a crosscutting aspect as the calculation of record was not reflective of current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:      Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain and Adequate Flooding Analysis The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after AmerenUE failed to provide adequate design control measures for verifying the adequacy of flooding analysis for the auxiliary feedwater pipe chase room 1206/1207. The revised calculation, performed on December 4, 2001, determined that the 10-inch piping from the condensate storage tank going to the main condenser was the limiting source of potential flooding. However several missing or incorrect assumptions challenged the basis for operability of safety related auxiliary feedwater pump transmitters located in the room 22 inches above the floor level. On December 16, 2009, the licensee reperformed the flooding analysis calculation, M-FL-04, Revision 5, including the main condenser as an additional source of flooding. Although 984 gpm of margin was lost due to inclusion of the condenser as a source, the revised analysis supported an operability determination for the transmitters as operable.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time, and did not increase the likelihood of a seismic, flooding, or severe weather initiating event. This finding was determined to not have a crosscutting aspect as the calculation of record was not reflective of current licensee performance.
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of Failure to Follow Operability Determination Procedure The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for two examples of failure to follow Procedure APA-ZZ-00500, Appendix 1, Operability and Functionality Determinations. The first example occurred on January 14, 2009, following an immediate operability determination made in response to Callaway Action Request 200900231. That Callaway action request documented significant emergency diesel generator heat exchanger tube wall thinning during eddy current testing. The operability determination performed in response to the degraded condition identified in Callaway Action Request 200900231 assumed a linear rate of degradation based on the rate observed from 2006 to 2008 and extrapolated forward to predict when heat exchanger tube plugging limits would be exceeded. Subsequent eddy current testing by the licensee found that the assumed linear degradation rate was nonconservative. The inspectors determined that the licensee failed to provide a reasonable expectation of operability consistent with the requirements of licensee Procedure APA-ZZ-00500, Appendix 1. Specifically, the licensee assumed a nonconservative linear rate of degradation for demonstrating emergency diesel heat exchanger operability despite empirical data that suggested the rate increased as a function of time.
The second example occurred on December 10, 2009, following initiation of Callaway Action Request 200910153 which documented that the steam generator C atmospheric steam dump valve (ABPV0003) would not repeatedly stroke to the same position. The Callaway action request documented that some amount of foreign material within the valve positioner was the cause of the repeatability issue with the valve. The inspectors reviewed Callaway Action Request 200910153 and noted that an immediate operability determination was not made on the identified degraded condition of foreign material within the air supply to the steam generator atmospheric steam dump valves. Since all four steam generator atmospheric steam dump valves share a common instrument air supply, the inspectors
 
determined that the licensee failed to identify what structures, systems, and components were affected by the degraded condition in Callaway Action Request 200910153. Following questioning by the inspectors, the licensee tested the remaining three steam generator atmospheric steam dump valves. During that testing, the licensee found the steam generator B atmospheric steam dump valve would not consistently stroke and that there was a small amount of foreign material within the air operated valve positioner.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations [H.1(b)].
Inspection Report# : 2009005 (pdf)
Barrier Integrity Significance:      Sep 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Adequate Administration Controls for Failed Containment Isolation Valve The inspectors identified a green noncited violation of Technical Specification 3.6.3, "Containment Isolation Valves,"
after the licensee failed to implement adequate administrative controls following the failure of valve EGHV0059. On August 10, 2010, containment isolation valve EGHV0059 failed to indicate full closed in the control room. The licensee declared the valve inoperable and isolated the affected penetration flow path. To ensure reactor coolant pump cooling the licensee unisolated the penetration by opening valve EGHV0131 and placing it under administrative controls. The on-shift operations technician was assigned to isolate the penetration in the event containment isolation was required. The resident inspectors found the licensees administrative controls were not consistent with the requirements in the technical specification bases which required a dedicated operator at the valve. The licensee then stationed a dedicated operator at valve EGHV0131 while repairs were conducted on valve EGHV0059. This issue was entered into the licensees corrective action program as Callaway Action Request 201007644.
This finding is more than minor because it was associated with the Barrier Integrity Cornerstone attribute of procedural quality and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to represent an actual open pathway in the physical integrity of reactor containment. Using Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," the issue was determined to be a Type B finding of very low safety significance since the containment penetration was associated with a closed system and would generally not contribute to large early release frequency. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure procedures used for addressing administrative controls were accurate and consistent with the technical specification bases [H.2(c)].
Inspection Report# : 2010004 (pdf)
Significance:      Jun 23, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Maintain Two Operable Source Range Channels During Core Alterations The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, when the licensees inadequate procedure and failure to control work activities during a reload of the reactor vessel fuel
 
assemblies resulted in deenergization of all available source range nuclear instrument channels. On May 6, 2010, while in Mode 6 - Refueling, licensee testing of nuclear instrument power range channel N44 and maintenance on 120 Vac instrument bus NN03 affecting power range channel N43 made up the logic for permissive P-10. The permissive sent a protective logic signal to deenergize both available source range nuclear instruments. The control room immediately directed the fuel handling crew to stop fuel movement until the source range channels could be restored. A fuel assembly was in the upender ready for transfer to the reactor vessel core location at the time. The licensee placed this issue into the corrective action program as Callaway Action Request 201004301.
This finding is more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or releases. Using Manual Chapter 0609 Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 - Operational Checklists for Both PWRs and BWRs, this finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory, did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and did not degrade the licensees ability to recover decay heat removal once lost. This finding had a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to coordinate work activities by incorporating actions to address the impact of the work on different job activities and communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2010003 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : November 29, 2010
 
Callaway 4Q/2010 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2010 Identified By: NRC Item Type: FIN Finding Failure to Ensure Completion of Corrective Actions for Degraded Chemical and Volume Control System Valves The inspectors identified a finding associated with AmerenUEs failure to take prompt corrective actions for leaking boundary valves in the chemical and volume control system. On April 13, 2010, an attempt to place the train A chemical and volume control system mixed bed in service resulted in leakage past a documented leaking drain valve.
The lingering equipment problems resulted in an unplanned 25 gallon per minute loss rate of volume control tank inventory and an emergency action level declaration for excessive reactor coolant system leakage. Later, the declaration was retracted. The licensee placed this issue into the corrective action program as Callaway Action Request 201003146.
This finding is more than minor because it was associated with the reactor safety Initiating Events Cornerstone attribute of configuration control and affected the objective to limit the likelihood of events that upset plant stability.
Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding is of very low significance because the condition did not result in the reactor coolant system technical specification leakage limit being exceeded, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or internal/external flood. This finding, which involved inadequate scheduling of corrective action related jobs, has a crosscutting aspect in the area of human performance associated with the work control component because AmerenUE did not appropriately coordinate work activities to address the impact of the work on different job activities [H.3(b)].
Inspection Report# : 2010003 (pdf)
Mitigating Systems Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Operability Determination Procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to follow Procedure APA ZZ 00500, Appendix 1, Operability and Functionality Determinations. On the morning of September 23, 2010, Callaway engineering was informed that a concern existed that the safety related portion of the component cooling water system safety function could be affected by a guillotine break at the nonsafety/nonseismic boundary for supply and return piping to the radwaste building. The inspectors determined that the licensee staff did not engage the shift manager early enough and the shift manager did not adequately challenge the basis describing the nonconforming condition as acceptable. The shift manager allowed the component cooling water system to be in an indeterminate state of operability for over two hours without putting compensatory measures in place as described in Procedure APA ZZ 00500, Appendix 1. This issue was entered into the licensees corrective action program as Callaway Action Request 201010739.
This finding was determined to be greater than minor because it impacted the mitigating systems cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of
 
systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as requiring a Phase 3 analysis. The NRC senior risk analyst determined that because ?CDF was less than 1E-6 and ?LERF was not a significant contributor to risk, this finding was of very low safety significance, Green. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations.
Inspection Report# : 2010005 (pdf)
Significance:      Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate, Untimely Corrective Actions for a Containment Spray System Condition Adverse to Quality The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
associated with the licensees failure to promptly identify and correct a boric acid leak on the containment spray system, a condition adverse to quality. During a plant walkdown on October 14, 2010, the inspectors noted the continued existence of a boric acid leak on the flow element above the discharge of the train A containment spray pump. Further inspection revealed the leak was first identified on February 16, 2009. The inspectors found that nearly twenty months after initial identification, the repair plan for the leak had not been assigned a scheduled date. The failure to promptly correct the leak was directly caused by a lack of coordination between the engineering and outage planning departments. This issue was entered into the licensees corrective action program as Callaway Action Request 201010263. Immediate corrective action included scheduling the repair for January 2011.
This finding is more than minor because, if left uncorrected, programmatic work control and corrective action deficiencies would have the potential to lead to a more significant safety concern. This finding affected the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded condition did not result in a loss of operability or functionality. The inspectors determined that the finding has a crosscutting aspect in the area of human performance because the licensee work practices did not ensure supervisory and management oversight of work activities, such that nuclear safety was supported.
Inspection Report# : 2010005 (pdf)
Significance:      Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Enter Condition Adverse to Quality Associated with Emergency Diesel Generator Jacket Water Keep Warm Pump into the Corrective Action Program The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to follow the requirements of Callaway Procedure APA ZZ 00500, Corrective Action Program, associated with a degraded train B emergency diesel generator jacket water keep warm pump. On November 6, 2010, the supply breaker to the train B emergency diesel generator jacket water keep warm pump tripped unexpectedly causing the engine to become inoperable. During follow-up investigation, the inspectors found that a March 31, 2009 motor circuit evaluation was performed that showed a step decrease in insulation resistance from 10,250 Mega-ohms to 3.5 Mega-ohms. The degradation was at a sufficient rate such that there was a reasonable doubt the motor would continue to be reliable until the next performance of the motor circuit evaluation. The licensee failed to recognize this degradation and, as a result, did not initiate a Callaway action request to evaluate the condition. This issue was entered into the licensees corrective action program as Callaway Action Request 201010654.
This finding is greater than minor because if left uncorrected, the failure to fully utilize the corrective action program could become a more significant safety concern. The inspectors determined that this finding impacted the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as having very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather
 
initiating events. The cause of this finding is related to the problem identification and resolution crosscutting component of the corrective action program because licensee personnel failed to implement a corrective action program with a low threshold for identifying issues.
Inspection Report# : 2010005 (pdf)
Significance:      Nov 05, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Degraded Conditions in Essential Service Water System in a Timely Manner The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to correct in a timely manner degraded conditions affecting the essential service water system. Specifically, the licensee failed to resolve the combined effects of corrosion and waterhammer events resulting in system leaks. The licensee has experienced the waterhammer events since initial plant startup and has been experiencing problems with corrosion since the mid 1990s. As corrective actions for this issue, the licensee plans to implement two system modifications next refueling outage to mitigate the impacts of waterhammer events. This noncited violation was entered into the corrective action program as Callaway Action Request 201010635.
The issue was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of human performance associated with the component of resources because the licensee did not maintain the plant to minimize long-standing equipment issues.
Inspection Report# : 2010006 (pdf)
Significance:      Nov 05, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Repetitive Failures in Steam Generator Atmospheric Dump Valves in a Timely Manner The team identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving the failure to promptly correct deficiencies affecting the steam generator atmospheric steam dump valves. In 2002, system engineers identified that the valves current-to-pressure transducers were experiencing degradation because they were subjected to high vibration, and a proposed modification to move the transducers to a low vibration area occurred in 2006. The licensee experienced several additional failures in 2009 and determined that the reliable life of the transducers was 18 months in the high vibration areas. As of the date of the inspection, only one transducer of the four had been moved to a low vibration location, and the team determined that corrective actions for this condition adverse to quality have not been timely. The licensee plans to implement modifications to relocate the remaining three transducers to a lower vibration environment in 2011. The issue was entered into the licensees corrective action program as Callaway Action Request 200910153.
This issue was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors found that even though the steam generator atmospheric steam dump valves were not able to meet their technical specification surveillance requirements of achieving the full open position the valves would open sufficiently to meet its intended safety function. Therefore, the issue was of very low safety significance since it was a design or qualification deficiency confirmed not to result in a loss of functionality. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to maintain long term plant safety by minimization of long-standing equipment issues associated with steam generator atmospheric steam dump valve current-to-pressure transducers.
 
Inspection Report# : 2010006 (pdf)
Significance:        Nov 05, 2010 Identified By: NRC Item Type: FIN Finding Failure to Follow the Corrective Action Program Procedure The team identified a finding involving the licensees failure to follow the corrective action program procedure for assigning significance levels to Callaway action requests. This deficiency resulted in the licensees failure to adequately evaluate the cause and extent of condition for a number of issues, and in some examples resulted in recurrences of the issues. In one example the licensee identified a jacket water leak on Emergency Diesel Generator B in 2008. This significant condition adverse to quality was assigned a Significance Level 3 which only required a lower tier cause evaluation, when the procedure identified a significant condition adverse to quality as an example of a Significance Level 1. The team identified additional examples involving degraded safety-related equipment and security-related issues. As corrective action, the licensee entered the issue into its corrective action program as Callaway Action Request 201010472.
This issue was determined to be greater than minor because if left uncorrected, the issue could become a more significant safety concern. The inspectors determined that the issue involving Callaway Action Request 200812985, the failure of emergency diesel generator train B due to a leak in the jacket water system, was of very low safety significance because it was bounded by the significance of NCV 05000483/2009007-01, Failure to Ensure Suitable Replacement Parts Essential for Emergency Diesel Generator Train B.
The team evaluated the issue involving Callaway Action Request 200810379, the failure of engineered safety feature power supply SA036E, using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. This issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events.
The team also evaluated several security-related examples of this finding that are described in Enclosure 2 of this letter. These security issues were also determined to be of very low security significance. Based on the sensitivity of security issues, Enclosure 2 is not publicly available because it contains security-related information.
This finding has a crosscutting aspect in the area of human performance associated with the component of training because training was needed for the screening committee to better understand a significant condition adverse to quality and to better understand the significance of security issues.
Inspection Report# : 2010006 (pdf)
Significance: SL-IV Sep 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Accurately Report a Condition that Could Have Prevented Fulfillment of a Safety Function The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), "Licensee Event Report System," for failure to report simultaneous inoperability of two steam generator atmospheric steam dump valves as a condition that could have prevented fulfillment of a safety function. On February 8, 2010, AmerenUE submitted Licensee Event Report 05000483/2009-005-00 to document that steam generator atmospheric steam dump valve ABPV0002 was out of service longer than allowed by Technical Specification 3.7.4, "Atmospheric Steam Dump Valves (ASDs)." The licensee event report also documented a period where valve ABPV0002 inoperability overlapped the inoperability of steam generator atmospheric steam dump valve ABPV0003. Callaway Final Safety Analysis Report Section 15.6.3.2.2.p. stated that all three intact steam generator atmospheric steam dump valves are credited in the cool down for a steam generator tube rupture. The inspectors determined that the licensee failed to adequately evaluate the reportability of having simultaneous inoperability of two steam generator atmospheric steam dump valves as a safety system functional failure. This issue was entered into the licensees corrective action program as Callaway Action Request 201006086 and on September 29, 2010, the licensee submitted Licensee Event Report 05000483/2009-005-001 to correct the reporting error.
 
This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was strictly associated with a traditional enforcement violation.
Inspection Report# : 2010004 (pdf)
Significance:      Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Lack of Maintenance as Cause of Diesel Generator Failure The inspectors identified a Green noncited violation of 10 Part 50, Appendix B, Criterion V, for the failure to accomplish a root cause evaluation in accordance with station procedures. Specifically, the licensee failed to identify and document that implementing Fairbanks Morse Owners Group recommended maintenance would have had a high likelihood of preventing the March 30, 2010, emergency diesel generator failure. As a result, the licensee did not classify the addition of maintenance on the governor and the governor drive as a corrective action, and the lack of maintenance was not evaluated for extent of condition and corrective actions, as applicable. This issue has been entered into the licensees corrective action program as Callaway Action Request 201008405.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events. Specifically, the evaluation failed to discover the lack of maintenance on the diesel governor and drive and the licensee failed to classify the maintenance as necessary. In addition, there was a potential for other recommended maintenance not being performed on mitigating equipment due to not evaluating the extent of condition and cause. Using NRC Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance because the finding did not result in the loss of safety function for mitigating equipment. This finding has a crosscutting aspect in the problem identification and resolution area associated with the operating experience component, in that the licensee failed to evaluate operating experience applicable to the root cause in a systematic and timely manner.
Inspection Report# : 2010007 (pdf)
Significance:      Jun 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Surveillance Procedure to Verify and Maintain Emergency Core Cooling System Operable The inspectors identified a noncited violation of Technical Specification 3.5.2, Emergency Core Cooling Systems.
Specifically Technical Specifications Surveillance Requirement 3.5.2.3, Verify the ECCS piping is full of water, was not being met by licensee Procedure OSP-SA-00003, Emergency Core Cooling System Flow Path Verification and Venting. On April 22, 2010, the inspectors discovered that the train B residual heat removal system discharge line EJ-024-ECB-10' did not have an accessible high point vent. The line was required by Callaway procedures to be either monitored by venting or tested using an ultrasonic method as described in the procedures acceptance criteria.
Callaway had identified the need to install a vent valve in line EJ-024-ECB-10' per modification MP-08-0016 prior to Refueling Outage 17. The licensee originally scheduled the vent valve installation during Refueling Outage 17, but had inappropriately deferred the maintenance to the next outage in fall 2011. As immediate corrective action, the licensee installed the vent valves in Refueling Outage 17 and placed this issue into the corrective action program as Callaway Action Request 201004078.
This finding is more than minor because it affected the Mitigating Systems Cornerstone procedure quality attribute and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding is of very low significance because it was only a design or qualification deficiency confirmed not to result in loss of operability. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the
 
licensee failed to use conservative assumptions in decision making and did not adopt a requirement to demonstrate that either venting or ultrasonic testing was needed to verify line EJ-024-ECB-10 was full of water [H.1(b)].
Inspection Report# : 2010003 (pdf)
Significance:        Jun 23, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Fabricate Replacement Gasket for Emergency Diesel Generator TrainA The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to adequately select suitable replacement gaskets essential to the operation of emergency diesel generator train A. On March 30, 2010, during performance of Procedure OSP-NE-00024A, Standby Diesel Generator A 24-Hour Run and Hot Restart Test, the emergency diesel generator train A unexpectedly lost speed and tripped after 16.7 hours of operation. Posttrip indications revealed that the diesel generator tripped from a stripped splined shaft in the governor drive housing. The failure of the splined shaft was caused by an improperly cut gasket which did not have the required oil port hole to allow proper lubrication of the drive assembly. The licensee replaced the damaged shaft and placed this issue in their corrective action program as Callaway Action Request 201002675.
This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The resident inspectors performed the initial significance determination for the diesel gasket finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved the loss of one train of safety related equipment for greater than its technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination using the pre-solved worksheet from the Risk Informed Inspection Notebook for Callaway Nuclear Generating Station, Revision 2.01a. The analyst assumed an exposure period of one year. The finding was potentially Yellow, which warranted further review. The senior reactor analyst subsequently performed a bounding Phase 3 significance determination and found the finding to be of very low safety significance (Green). The dominant cutsets included a loss of offsite power initiating event, failure to recover offsite power in 4 hours, failure of the train B emergency diesel generator, and a reactor coolant pump seal failure. Equipment that mitigated the significance included the operable emergency diesel generator and the turbine-driven auxiliary feedwater pump. This finding did not have a crosscutting aspect since it was not a performance deficiency reflective of current licensee performance.
Inspection Report# : 2010003 (pdf)
Significance:        Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Operability Determination Procedure The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to follow Procedure APA-ZZ-00500, Appendix 1, Operability and Functionality Determinations. The inspectors determined that the licensee failed to provide a reasonable expectation of operability for the degraded condition. Specifically, the licensee failed to account for both auxiliary feedwater as an essential service water system load and fouling resistance in the component cooling water system heat exchanger. Long term corrective actions planned include a modification of the component cooling water heat exchangers divider plate during the upcoming April 2010 refueling outage. The licensee placed this issue in their corrective action program as Callaway Action Request 201001152.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification
 
allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations [H.1(b)].
Inspection Report# : 2010002 (pdf)
Significance:        Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Suitable Replacement Parts Essential for the Operation of the Component Cooling Water System The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to adequately select suitable replacement gaskets essential to the operation of the component cooling water system heat exchangers. On October 19, 2008, Callaway engineering personnel identified that the component cooling water heat exchangers, due to corrosion and inadequate gasket sealing, had a small gap between the divider plate and channel head such that it allowed essential service water flow to bypass the heat exchanger which resulted in a reduced heat transfer capability. Corrective actions to address the identified gap in the component cooling water heat exchanger were scheduled to be implemented during the licensees next refueling outage. The licensee entered the issue in the corrective action program as Callaway Action Request 201001900.
This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. This finding was determined not to have a crosscutting aspect since it is a performance deficiency not reflective of current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:        Mar 24, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to M aintain an Adequare Ultimate Heat Sink Thermal Performance Analysis The NRC identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after AmerenUE failed to provide adequate design control measures for verifying the adequacy of the ultimate heat sink thermal performance analysis evaluating the impact of heat rejected during a large break loss of coolant accident. The thermal performance analysis, most recently revised in 2007, did not account for a potential single active failure of each trains motor-operated valve designed to redirect the essential service water return flow up and over the tower fill material. With further analysis the licensee determined that a compensatory measure implementing a more restrictive initial operating range based on pond volume and initial temperature would ensure that the ultimate heat sink pond will not exceed its maximum temperature of 92.3 degrees Fahrenheit during a design basis accident. Corrective actions were being developed using Callaway Action Request 201001813.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A resident inspector performed the initial significance determination for the inoperable essential service water system, under certain conditions, using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved the potential inoperability of both trains of essential service water for greater than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and found that the finding was potentially greater than green. The senior reactor analyst then performed a bounding Phase 3 significance determination and found the finding to be of very low safety significance (Green).
The dominant core damage sequences included a medium break loss of coolant accident concurrent with the failure of essential service water system cooling tower bypass valves. The finding was mitigated because the motor operated
 
valves remained functional throughout the year, which minimized the frequencies for the scenarios of interest. This finding was determined to not have a crosscutting aspect as the calculation of record was not reflective of current licensee performance.
Inspection Report# : 2010002 (pdf)
Barrier Integrity Significance:      Sep 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Adequate Administration Controls for Failed Containment Isolation Valve The inspectors identified a green noncited violation of Technical Specification 3.6.3, "Containment Isolation Valves,"
after the licensee failed to implement adequate administrative controls following the failure of valve EGHV0059. On August 10, 2010, containment isolation valve EGHV0059 failed to indicate full closed in the control room. The licensee declared the valve inoperable and isolated the affected penetration flow path. To ensure reactor coolant pump cooling the licensee unisolated the penetration by opening valve EGHV0131 and placing it under administrative controls. The on-shift operations technician was assigned to isolate the penetration in the event containment isolation was required. The resident inspectors found the licensees administrative controls were not consistent with the requirements in the technical specification bases which required a dedicated operator at the valve. The licensee then stationed a dedicated operator at valve EGHV0131 while repairs were conducted on valve EGHV0059. This issue was entered into the licensees corrective action program as Callaway Action Request 201007644.
This finding is more than minor because it was associated with the Barrier Integrity Cornerstone attribute of procedural quality and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to represent an actual open pathway in the physical integrity of reactor containment. Using Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," the issue was determined to be a Type B finding of very low safety significance since the containment penetration was associated with a closed system and would generally not contribute to large early release frequency. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure procedures used for addressing administrative controls were accurate and consistent with the technical specification bases [H.2(c)].
Inspection Report# : 2010004 (pdf)
Significance:      Jun 23, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Maintain Two Operable Source Range Channels During Core Alterations The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, when the licensees inadequate procedure and failure to control work activities during a reload of the reactor vessel fuel assemblies resulted in deenergization of all available source range nuclear instrument channels. On May 6, 2010, while in Mode 6 - Refueling, licensee testing of nuclear instrument power range channel N44 and maintenance on 120 Vac instrument bus NN03 affecting power range channel N43 made up the logic for permissive P-10. The permissive sent a protective logic signal to deenergize both available source range nuclear instruments. The control room immediately directed the fuel handling crew to stop fuel movement until the source range channels could be restored. A fuel assembly was in the upender ready for transfer to the reactor vessel core location at the time. The licensee placed this issue into the corrective action program as Callaway Action Request 201004301.
This finding is more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or releases. Using Manual Chapter 0609 Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 - Operational
 
Checklists for Both PWRs and BWRs, this finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory, did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and did not degrade the licensees ability to recover decay heat removal once lost. This finding had a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to coordinate work activities by incorporating actions to address the impact of the work on different job activities and communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2010003 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Nov 05, 2010 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The team concluded that the corrective action program at the Callaway Plant was performing in a satisfactory manner to ensure safe plant operations. However, the team identified a number of instances in which the licensee did not follow its procedural guidance for assigning significance levels to problems identified and, as a result, did not adequately evaluate the causes and/or extent of conditions resulting in several repetitive issues.
The inspectors determined that the licensee evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. The inspectors noted that operating experience was considered in cause evaluations.
The team determined that the licensee had a healthy safety-conscious work environment in that workers felt free to raise safety concerns without fear of retaliation using all avenues available.
Inspection Report# : 2010006 (pdf)
Last modified : March 03, 2011
 
Callaway 1Q/2011 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2010 Identified By: NRC Item Type: FIN Finding Failure to Ensure Completion of Corrective Actions for Degraded Chemical and Volume Control System Valves The inspectors identified a finding associated with AmerenUEs failure to take prompt corrective actions for leaking boundary valves in the chemical and volume control system. On April 13, 2010, an attempt to place the train A chemical and volume control system mixed bed in service resulted in leakage past a documented leaking drain valve.
The lingering equipment problems resulted in an unplanned 25 gallon per minute loss rate of volume control tank inventory and an emergency action level declaration for excessive reactor coolant system leakage. Later, the declaration was retracted. The licensee placed this issue into the corrective action program as Callaway Action Request 201003146.
This finding is more than minor because it was associated with the reactor safety Initiating Events Cornerstone attribute of configuration control and affected the objective to limit the likelihood of events that upset plant stability.
Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding is of very low significance because the condition did not result in the reactor coolant system technical specification leakage limit being exceeded, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would be unavailable, and did not increase the likelihood of a fire or internal/external flood. This finding, which involved inadequate scheduling of corrective action related jobs, has a crosscutting aspect in the area of human performance associated with the work control component because AmerenUE did not appropriately coordinate work activities to address the impact of the work on different job activities [H.3(b)].
Inspection Report# : 2010003 (pdf)
Mitigating Systems Significance:        Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Document Reasonable Expectation of Operability for Equipment Supported b y the Class 1E Air Conditioning Units The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to adequately evaluate past operability associated with the Class 1E electrical equipment air conditioning unit. The inspectors identified that Revision 1 and 2 to Callaway Action Request 200800615 incorrectly concluded that the equipment supported by the Class 1E electrical equipment air conditioning unit train B was operable with the units cooling water flow control valve in manual. This issue was entered into the licensees corrective action program as Callaway Action Request 201102565.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened to a Phase 2 significance determination because it involved the loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a bounding Phase 3
 
significance determination and determined that the finding was of very low safety significance (Green). The very short exposure period coupled with the availability of train A equipment helped to mitigate the significance. The dominant core damage sequences included a loss of main feedwater initiating event; the loss of train B electrical power; and various failures of auxiliary feedwater. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions including verifying the validity of the underlying assumptions when performing operability/reportability evaluations.
Inspection Report# : 2011002 (pdf)
Significance: SL-IV Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report Inoperability of Class IE Electrical Equipment for a Period Greater than Allowed by the Plant's Technical Specifications The inspectors identified a IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for failure to report inoperability of Class 1E electrical equipment for a period greater than allowed by the plants technical specifications. The licensee determined there were no prior instances where the Class 1E electrical equipment air conditioning units were inoperable greater than the technical specification allowed completion time of the supported equipment. The inspectors reviewed the licensees reportability evaluation and identified that the event described in Callaway Action Request 200800615 resulted in a period where the Class 1E electrical equipment air conditioning unit train B was inoperable for approximately 37 hours which exceeded the technical specification allowed completion time of the equipment supported by the Class 1E electrical equipment and constituted a condition which was prohibited by the plant's technical specifications and should have been reported in a licensee event report.
This issue was entered into the licensees corrective action program as Callaway Action Request 201011132.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because in order to perform its regulatory function, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was strictly associated with a traditional enforcement violation.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Containment Spray Test Procedure Potentially Creates an Unanalyzed Condition The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to provide adequate procedural guidance for testing of containment spray pumps. The inspectors reviewed a licensee evaluation of the acceptability of their existing containment spray pump testing procedure and found that it failed to adequately address the underlying technical issues because it relied on operators recognizing the diversion flow path and focused on the operability of the containment spray system and not the ability to maintain the long term cooling function of the emergency core cooling system. Additionally, the inspectors identified that the procedure would have provided a diversion flow path of post-accident sump fluids back to the refueling water storage tank exceeding those currently analyzed in the Callaway licensing bases. This issue was entered into the licensees corrective action program as Callaway Action Request 201011233 and the licensee implemented procedure changes to address the potential for post-loss of coolant accident containment sump fluids being injected back to the refueling water storage tank.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of procedure quality and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding screened to a Phase 2 significance determination because it involved a potential loss of safety function. A Region IV senior reactor analyst performed a bounding Phase 3 significance determination and determined that the finding was of very low safety significance (Green). The very short exposure period coupled with the availability of equipment needed for other initiating events (other than small and medium loss of coolant accidents) helped to
 
mitigate the significance. The dominant core damage sequences included small and medium break loss of coolant accidents, and the failure of emergency core cooling pumps in the recirculation mode. This finding was determined not to have a crosscutting aspect since the performance deficiency is not reflective of current performance.
Inspection Report# : 2011002 (pdf)
Significance:      Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding Installation Inadequacy The inspectors identified a noncited violation of Technical Specification 5.4.1.a for failure to properly implement Procedure MDP-ZZ-S0001, "Scaffolding Installation and Evaluation," Revision 26, when scaffolding was erected near or in contact with equipment in safety-related structures. On February 8 and March 16, 2011, the inspectors identified two locations where scaffold poles and a scaffold pin were less than the procedure required 1 inch from the auxiliary building vent line, the Train B emergency diesel lube oil drain line, and also essential service water system piping in the Train B diesel room. This issue was entered into the licensees corrective action program as Callaway Action Request 201102091.
The deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding was associated with the Mitigating Systems Cornerstone. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the component of corrective action program because the licensee did not have a low threshold for identifying scaffold issues.
Inspection Report# : 2011002 (pdf)
Significance: SL-IV Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report The team identified a Severity Level IV, noncited violation of 10 CFR 50.71, Maintenance of records, making of reports, paragraph (e) which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the updated safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Specifically, the licensee incorporated numerous errors in the updated safety analysis report associated with the descriptions of the onsite electrical power systems. The licensee has entered this violation into their corrective action program as Condition Reports 201101335 and 201102064.
The inspectors determined that the failure to update the updated safety analysis report as required by 10 CFR 50.71(e),
Maintenance of records, making of reports was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function.
The inspectors used the NRC Enforcement Policy, dated September 30, 2010, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was associated with a traditional enforcement violation.
Inspection Report# : 2011006 (pdf)
Significance:      Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Design the Emergency Diesel Generator Ground Fault Protection Circuitry
 
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, when designing the bypass circuitry for the emergency diesel generator ground fault trip function, the licensee failed to ensure that the associated electrical components were adequately designed for the continuous duty they would have to withstand under bypassed trip conditions. This could result in an ignition source and subsequent fire in the area under these conditions. This finding was entered into the licensees corrective action program as Condition Report 201102064.
The team determined that the failure to analyze the suitability of the emergency diesel generator components when protection features were bypassed was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design of these components could have prevented continued operation of the emergency diesel generator under ground fault conditions with the trip signal bypassed. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Specifically, the licensee revised the associated procedures to include these components in the combustible material exclusion zone. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Residual Heat Removal Flow Alarm Setpoint The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of March 3, 2011, the Mode 6 residual heat removal system low flow alarm setpoint did not adequately account for flow measurement uncertainties, and consequently was non-conservative. The licensee has entered the violation into their corrective action program as Condition Report 201101750.
The team determined that the failure to adequately analyze the uncertainty in measurement of residual heat removal system flow, and the impact of this failure, was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design basis analysis, and plant instrumentation, did not ensure that, while operating in Mode 6, the control room operators would be alerted whenever the residual heat removal system flow through the reactor coolant system was below the required value of 1000 gallons per minute. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011006 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Operability Determination Procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to follow Procedure APA ZZ 00500, Appendix 1, Operability and Functionality Determinations. On the morning of September 23, 2010, Callaway engineering was informed that a concern existed
 
that the safety related portion of the component cooling water system safety function could be affected by a guillotine break at the nonsafety/nonseismic boundary for supply and return piping to the radwaste building. The inspectors determined that the licensee staff did not engage the shift manager early enough and the shift manager did not adequately challenge the basis describing the nonconforming condition as acceptable. The shift manager allowed the component cooling water system to be in an indeterminate state of operability for over two hours without putting compensatory measures in place as described in Procedure APA ZZ 00500, Appendix 1. This issue was entered into the licensees corrective action program as Callaway Action Request 201010739.
This finding was determined to be greater than minor because it impacted the mitigating systems cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as requiring a Phase 3 analysis. The NRC senior risk analyst determined that because ?CDF was less than 1E-6 and ?LERF was not a significant contributor to risk, this finding was of very low safety significance, Green. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations.
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate, Untimely Corrective Actions for a Containment Spray System Condition Adverse to Quality The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
associated with the licensees failure to promptly identify and correct a boric acid leak on the containment spray system, a condition adverse to quality. During a plant walkdown on October 14, 2010, the inspectors noted the continued existence of a boric acid leak on the flow element above the discharge of the train A containment spray pump. Further inspection revealed the leak was first identified on February 16, 2009. The inspectors found that nearly twenty months after initial identification, the repair plan for the leak had not been assigned a scheduled date. The failure to promptly correct the leak was directly caused by a lack of coordination between the engineering and outage planning departments. This issue was entered into the licensees corrective action program as Callaway Action Request 201010263. Immediate corrective action included scheduling the repair for January 2011.
This finding is more than minor because, if left uncorrected, programmatic work control and corrective action deficiencies would have the potential to lead to a more significant safety concern. This finding affected the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded condition did not result in a loss of operability or functionality. The inspectors determined that the finding has a crosscutting aspect in the area of human performance because the licensee work practices did not ensure supervisory and management oversight of work activities, such that nuclear safety was supported.
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Enter Condition Adverse to Quality Associated with Emergency Diesel Generator Jacket Water Keep Warm Pump into the Corrective Action Program The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to follow the requirements of Callaway Procedure APA ZZ 00500, Corrective Action Program, associated with a degraded train B emergency diesel generator jacket water keep warm pump. On November 6, 2010, the supply breaker to the train B emergency diesel generator jacket water keep warm pump tripped unexpectedly causing the engine to become inoperable. During follow-up investigation, the inspectors found that a March 31, 2009 motor circuit evaluation was performed that showed a step decrease in insulation resistance from 10,250 Mega-ohms to 3.5 Mega-ohms. The degradation was at a sufficient rate such that there was a reasonable doubt the motor would continue to be reliable until the next performance of the motor circuit
 
evaluation. The licensee failed to recognize this degradation and, as a result, did not initiate a Callaway action request to evaluate the condition. This issue was entered into the licensees corrective action program as Callaway Action Request 201010654.
This finding is greater than minor because if left uncorrected, the failure to fully utilize the corrective action program could become a more significant safety concern. The inspectors determined that this finding impacted the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as having very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. The cause of this finding is related to the problem identification and resolution crosscutting component of the corrective action program because licensee personnel failed to implement a corrective action program with a low threshold for identifying issues.
Inspection Report# : 2010005 (pdf)
Significance:      Nov 05, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Degraded Conditions in Essential Service Water System in a Timely Manner The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to correct in a timely manner degraded conditions affecting the essential service water system. Specifically, the licensee failed to resolve the combined effects of corrosion and waterhammer events resulting in system leaks. The licensee has experienced the waterhammer events since initial plant startup and has been experiencing problems with corrosion since the mid 1990s. As corrective actions for this issue, the licensee plans to implement two system modifications next refueling outage to mitigate the impacts of waterhammer events. This noncited violation was entered into the corrective action program as Callaway Action Request 201010635.
The issue was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of human performance associated with the component of resources because the licensee did not maintain the plant to minimize long-standing equipment issues.
Inspection Report# : 2010006 (pdf)
Significance:      Nov 05, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Repetitive Failures in Steam Generator Atmospheric Dump Valves in a Timely Manner The team identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving the failure to promptly correct deficiencies affecting the steam generator atmospheric steam dump valves. In 2002, system engineers identified that the valves current-to-pressure transducers were experiencing degradation because they were subjected to high vibration, and a proposed modification to move the transducers to a low vibration area occurred in 2006. The licensee experienced several additional failures in 2009 and determined that the reliable life of the transducers was 18 months in the high vibration areas. As of the date of the inspection, only one transducer of the four had been moved to a low vibration location, and the team determined that corrective actions for this condition adverse to quality have not been timely. The licensee plans to implement modifications to relocate the remaining three transducers to a lower vibration environment in 2011. The issue was entered into the licensees corrective action program as Callaway Action Request 200910153.
 
This issue was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors found that even though the steam generator atmospheric steam dump valves were not able to meet their technical specification surveillance requirements of achieving the full open position the valves would open sufficiently to meet its intended safety function. Therefore, the issue was of very low safety significance since it was a design or qualification deficiency confirmed not to result in a loss of functionality. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to maintain long term plant safety by minimization of long-standing equipment issues associated with steam generator atmospheric steam dump valve current-to-pressure transducers.
Inspection Report# : 2010006 (pdf)
Significance:        Nov 05, 2010 Identified By: NRC Item Type: FIN Finding Failure to Follow the Corrective Action Program Procedure The team identified a finding involving the licensees failure to follow the corrective action program procedure for assigning significance levels to Callaway action requests. This deficiency resulted in the licensees failure to adequately evaluate the cause and extent of condition for a number of issues, and in some examples resulted in recurrences of the issues. In one example the licensee identified a jacket water leak on Emergency Diesel Generator B in 2008. This significant condition adverse to quality was assigned a Significance Level 3 which only required a lower tier cause evaluation, when the procedure identified a significant condition adverse to quality as an example of a Significance Level 1. The team identified additional examples involving degraded safety-related equipment and security-related issues. As corrective action, the licensee entered the issue into its corrective action program as Callaway Action Request 201010472.
This issue was determined to be greater than minor because if left uncorrected, the issue could become a more significant safety concern. The inspectors determined that the issue involving Callaway Action Request 200812985, the failure of emergency diesel generator train B due to a leak in the jacket water system, was of very low safety significance because it was bounded by the significance of NCV 05000483/2009007-01, Failure to Ensure Suitable Replacement Parts Essential for Emergency Diesel Generator Train B.
The team evaluated the issue involving Callaway Action Request 200810379, the failure of engineered safety feature power supply SA036E, using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. This issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events.
The team also evaluated several security-related examples of this finding that are described in Enclosure 2 of this letter. These security issues were also determined to be of very low security significance. Based on the sensitivity of security issues, Enclosure 2 is not publicly available because it contains security-related information.
This finding has a crosscutting aspect in the area of human performance associated with the component of training because training was needed for the screening committee to better understand a significant condition adverse to quality and to better understand the significance of security issues.
Inspection Report# : 2010006 (pdf)
Significance: SL-IV Sep 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Accurately Report a Condition that Could Have Prevented Fulfillment of a Safety Function The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), "Licensee Event Report System," for failure to report simultaneous inoperability of two steam generator atmospheric steam dump valves as a condition that could have prevented fulfillment of a safety function. On February 8, 2010, AmerenUE submitted
 
Licensee Event Report 05000483/2009-005-00 to document that steam generator atmospheric steam dump valve ABPV0002 was out of service longer than allowed by Technical Specification 3.7.4, "Atmospheric Steam Dump Valves (ASDs)." The licensee event report also documented a period where valve ABPV0002 inoperability overlapped the inoperability of steam generator atmospheric steam dump valve ABPV0003. Callaway Final Safety Analysis Report Section 15.6.3.2.2.p. stated that all three intact steam generator atmospheric steam dump valves are credited in the cool down for a steam generator tube rupture. The inspectors determined that the licensee failed to adequately evaluate the reportability of having simultaneous inoperability of two steam generator atmospheric steam dump valves as a safety system functional failure. This issue was entered into the licensees corrective action program as Callaway Action Request 201006086 and on September 29, 2010, the licensee submitted Licensee Event Report 05000483/2009-005-001 to correct the reporting error.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was strictly associated with a traditional enforcement violation.
Inspection Report# : 2010004 (pdf)
Significance:      Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Lack of Maintenance as Cause of Diesel Generator Failure The inspectors identified a Green noncited violation of 10 Part 50, Appendix B, Criterion V, for the failure to accomplish a root cause evaluation in accordance with station procedures. Specifically, the licensee failed to identify and document that implementing Fairbanks Morse Owners Group recommended maintenance would have had a high likelihood of preventing the March 30, 2010, emergency diesel generator failure. As a result, the licensee did not classify the addition of maintenance on the governor and the governor drive as a corrective action, and the lack of maintenance was not evaluated for extent of condition and corrective actions, as applicable. This issue has been entered into the licensees corrective action program as Callaway Action Request 201008405.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events. Specifically, the evaluation failed to discover the lack of maintenance on the diesel governor and drive and the licensee failed to classify the maintenance as necessary. In addition, there was a potential for other recommended maintenance not being performed on mitigating equipment due to not evaluating the extent of condition and cause. Using NRC Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance because the finding did not result in the loss of safety function for mitigating equipment. This finding has a crosscutting aspect in the problem identification and resolution area associated with the operating experience component, in that the licensee failed to evaluate operating experience applicable to the root cause in a systematic and timely manner.
Inspection Report# : 2010007 (pdf)
Significance:      Jun 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Surveillance Procedure to Verify and Maintain Emergency Core Cooling System Operable The inspectors identified a noncited violation of Technical Specification 3.5.2, Emergency Core Cooling Systems.
Specifically Technical Specifications Surveillance Requirement 3.5.2.3, Verify the ECCS piping is full of water, was not being met by licensee Procedure OSP-SA-00003, Emergency Core Cooling System Flow Path Verification and Venting. On April 22, 2010, the inspectors discovered that the train B residual heat removal system discharge line EJ-024-ECB-10' did not have an accessible high point vent. The line was required by Callaway procedures to be either monitored by venting or tested using an ultrasonic method as described in the procedures acceptance criteria.
Callaway had identified the need to install a vent valve in line EJ-024-ECB-10' per modification MP-08-0016 prior to Refueling Outage 17. The licensee originally scheduled the vent valve installation during Refueling Outage 17, but
 
had inappropriately deferred the maintenance to the next outage in fall 2011. As immediate corrective action, the licensee installed the vent valves in Refueling Outage 17 and placed this issue into the corrective action program as Callaway Action Request 201004078.
This finding is more than minor because it affected the Mitigating Systems Cornerstone procedure quality attribute and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that this finding is of very low significance because it was only a design or qualification deficiency confirmed not to result in loss of operability. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions in decision making and did not adopt a requirement to demonstrate that either venting or ultrasonic testing was needed to verify line EJ-024-ECB-10 was full of water [H.1(b)].
Inspection Report# : 2010003 (pdf)
Significance:        Jun 23, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Fabricate Replacement Gasket for Emergency Diesel Generator TrainA The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to adequately select suitable replacement gaskets essential to the operation of emergency diesel generator train A. On March 30, 2010, during performance of Procedure OSP-NE-00024A, Standby Diesel Generator A 24-Hour Run and Hot Restart Test, the emergency diesel generator train A unexpectedly lost speed and tripped after 16.7 hours of operation. Posttrip indications revealed that the diesel generator tripped from a stripped splined shaft in the governor drive housing. The failure of the splined shaft was caused by an improperly cut gasket which did not have the required oil port hole to allow proper lubrication of the drive assembly. The licensee replaced the damaged shaft and placed this issue in their corrective action program as Callaway Action Request 201002675.
This finding was greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The resident inspectors performed the initial significance determination for the diesel gasket finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved the loss of one train of safety related equipment for greater than its technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination using the pre-solved worksheet from the Risk Informed Inspection Notebook for Callaway Nuclear Generating Station, Revision 2.01a. The analyst assumed an exposure period of one year. The finding was potentially Yellow, which warranted further review. The senior reactor analyst subsequently performed a bounding Phase 3 significance determination and found the finding to be of very low safety significance (Green). The dominant cutsets included a loss of offsite power initiating event, failure to recover offsite power in 4 hours, failure of the train B emergency diesel generator, and a reactor coolant pump seal failure. Equipment that mitigated the significance included the operable emergency diesel generator and the turbine-driven auxiliary feedwater pump. This finding did not have a crosscutting aspect since it was not a performance deficiency reflective of current licensee performance.
Inspection Report# : 2010003 (pdf)
Barrier Integrity Significance:        Sep 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Adequate Administration Controls for Failed Containment Isolation Valve The inspectors identified a green noncited violation of Technical Specification 3.6.3, "Containment Isolation Valves,"
 
after the licensee failed to implement adequate administrative controls following the failure of valve EGHV0059. On August 10, 2010, containment isolation valve EGHV0059 failed to indicate full closed in the control room. The licensee declared the valve inoperable and isolated the affected penetration flow path. To ensure reactor coolant pump cooling the licensee unisolated the penetration by opening valve EGHV0131 and placing it under administrative controls. The on-shift operations technician was assigned to isolate the penetration in the event containment isolation was required. The resident inspectors found the licensees administrative controls were not consistent with the requirements in the technical specification bases which required a dedicated operator at the valve. The licensee then stationed a dedicated operator at valve EGHV0131 while repairs were conducted on valve EGHV0059. This issue was entered into the licensees corrective action program as Callaway Action Request 201007644.
This finding is more than minor because it was associated with the Barrier Integrity Cornerstone attribute of procedural quality and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to represent an actual open pathway in the physical integrity of reactor containment. Using Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," the issue was determined to be a Type B finding of very low safety significance since the containment penetration was associated with a closed system and would generally not contribute to large early release frequency. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure procedures used for addressing administrative controls were accurate and consistent with the technical specification bases [H.2(c)].
Inspection Report# : 2010004 (pdf)
Significance:      Jun 23, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Maintain Two Operable Source Range Channels During Core Alterations The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, when the licensees inadequate procedure and failure to control work activities during a reload of the reactor vessel fuel assemblies resulted in deenergization of all available source range nuclear instrument channels. On May 6, 2010, while in Mode 6 - Refueling, licensee testing of nuclear instrument power range channel N44 and maintenance on 120 Vac instrument bus NN03 affecting power range channel N43 made up the logic for permissive P-10. The permissive sent a protective logic signal to deenergize both available source range nuclear instruments. The control room immediately directed the fuel handling crew to stop fuel movement until the source range channels could be restored. A fuel assembly was in the upender ready for transfer to the reactor vessel core location at the time. The licensee placed this issue into the corrective action program as Callaway Action Request 201004301.
This finding is more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or releases. Using Manual Chapter 0609 Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 - Operational Checklists for Both PWRs and BWRs, this finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory, did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and did not degrade the licensees ability to recover decay heat removal once lost. This finding had a crosscutting aspect in the area of human performance associated with the work control component because the licensee failed to coordinate work activities by incorporating actions to address the impact of the work on different job activities and communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2010003 (pdf)
Emergency Preparedness
 
Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Nov 05, 2010 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The team concluded that the corrective action program at the Callaway Plant was performing in a satisfactory manner to ensure safe plant operations. However, the team identified a number of instances in which the licensee did not follow its procedural guidance for assigning significance levels to problems identified and, as a result, did not adequately evaluate the causes and/or extent of conditions resulting in several repetitive issues.
The inspectors determined that the licensee evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. The inspectors noted that operating experience was considered in cause evaluations.
The team determined that the licensee had a healthy safety-conscious work environment in that workers felt free to raise safety concerns without fear of retaliation using all avenues available.
Inspection Report# : 2010006 (pdf)
Last modified : June 07, 2011
 
Callaway 2Q/2011 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Implement a Plant Safety System Test Procedure A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified when the licensees failure to correctly follow a test procedure resulted in a negative reactivity excursion due to excessive boration. On May 27, 2011, with the Callaway Plant at 100 percent power, maintenance was in progress to perform a functional test of the plants safety system trip actuating devices. During the test the instrument maintenance technicians failed to place the mode selector switch in the test position. This resulted in switching the charging pump suction from the volume control tank to the refueling water storage tank. The inadvertent actuation resulted in a reactivity excursion that required lowering main turbine power and reactor power to about 92 percent. The crew stabilized the plant and returned critical parameters to their normal control bands. The licensee entered this issue in the corrective action program as Callaway Action Request 201104451.
This finding is more than minor because it was associated with the configuration control attribute of the Initiating Events Cornerstone and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions will not be available. This finding had a cross-cutting aspect in the area of human performance associated with the work practices component because the instrument maintenance technicians failed to adequately use human error prevention techniques, such as self- and peer-checking to ensure that work activities are performed safely Inspection Report# : 2011003 (pdf)
Mitigating Systems Significance:        Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Flooding Analysis for Room 3101 The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to provide adequate design control measures for verifying the adequacy of the flooding analysis associated with the 2009 modification that replaced essential service water carbon steel piping with high density polyethylene piping. The licensee did not update the flooding analysis of record to consider potential failures in the new piping. The licensee generated Callaway Action Request 201102957 to develop a means to evaluate the relative stresses associated with the new pipe.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding required a Phase 2 significance determination. Using the presolved worksheet from the Risk Informed Inspection Notebook for the Callaway Station, Revision 2.01a, the finding was red, which warranted further review. Therefore, a senior reactor analyst
 
performed a bounding Phase 3 significance determination. The bounding change to the core damage frequency was approximately 4.1E-7 (Green). This was impacted significantly by the very small amount of new piping in the room.
This finding was determined to have a cross-cutting aspect in the area of Problem Identification and Resolution associated with the corrective action component in that the licensee did not thoroughly evaluate the extent of condition when the residents challenged the flooding calculation in December 2010 such that the resolutions addressed causes and extent of conditions, as necessary Inspection Report# : 2011003 (pdf)
Significance:      Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Analyze Refueling Water Storage Tank Level Transmitters for High-Energy Line Break The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to adequately evaluate a potential high-energy line break in nonseismically qualified auxiliary steam piping in the refueling water storage tank valve house. The harsh environment from a high-energy line break had the potential to impact safety related level transmitters associated with the refueling water storage tank. Following identification of this issue by the inspectors, the licensee analyzed the nonnuclear auxiliary piping to ensure it could withstand safe shutdown earthquake loadings which allowed high-energy line breaks at intermediate locations to be excluded. This issue was entered into the licensees corrective action program as Callaway Action Request 201102588.
This finding is greater than minor because it is associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, this finding is determined to be of very low safety significance since subsequent evaluation concluded the issue was a design or qualification deficiency confirmed not to result in loss of operability or functionality. This finding did not have a cross-cutting aspect since the error associated with the high-energy line break analysis was not reflective of current licensee performance.
Inspection Report# : 2011003 (pdf)
Significance:      Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Document Reasonable Expectation of Operability for Equipment Supported b y the Class 1E Air Conditioning Units The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to adequately evaluate past operability associated with the Class 1E electrical equipment air conditioning unit. The inspectors identified that Revision 1 and 2 to Callaway Action Request 200800615 incorrectly concluded that the equipment supported by the Class 1E electrical equipment air conditioning unit train B was operable with the units cooling water flow control valve in manual. This issue was entered into the licensees corrective action program as Callaway Action Request 201102565.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened to a Phase 2 significance determination because it involved the loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a bounding Phase 3 significance determination and determined that the finding was of very low safety significance (Green). The very short exposure period coupled with the availability of train A equipment helped to mitigate the significance. The dominant core damage sequences included a loss of main feedwater initiating event; the loss of train B electrical power; and various failures of auxiliary feedwater. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions including verifying the validity of the underlying assumptions when performing operability/reportability evaluations.
 
Inspection Report# : 2011002 (pdf)
Significance: SL-IV Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report Inoperability of Class IE Electrical Equipment for a Period Greater than Allowed by the Plant's Technical Specifications The inspectors identified a IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for failure to report inoperability of Class 1E electrical equipment for a period greater than allowed by the plants technical specifications. The licensee determined there were no prior instances where the Class 1E electrical equipment air conditioning units were inoperable greater than the technical specification allowed completion time of the supported equipment. The inspectors reviewed the licensees reportability evaluation and identified that the event described in Callaway Action Request 200800615 resulted in a period where the Class 1E electrical equipment air conditioning unit train B was inoperable for approximately 37 hours which exceeded the technical specification allowed completion time of the equipment supported by the Class 1E electrical equipment and constituted a condition which was prohibited by the plant's technical specifications and should have been reported in a licensee event report.
This issue was entered into the licensees corrective action program as Callaway Action Request 201011132.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because in order to perform its regulatory function, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was strictly associated with a traditional enforcement violation.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Containment Spray Test Procedure Potentially Creates an Unanalyzed Condition The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to provide adequate procedural guidance for testing of containment spray pumps. The inspectors reviewed a licensee evaluation of the acceptability of their existing containment spray pump testing procedure and found that it failed to adequately address the underlying technical issues because it relied on operators recognizing the diversion flow path and focused on the operability of the containment spray system and not the ability to maintain the long term cooling function of the emergency core cooling system. Additionally, the inspectors identified that the procedure would have provided a diversion flow path of post-accident sump fluids back to the refueling water storage tank exceeding those currently analyzed in the Callaway licensing bases. This issue was entered into the licensees corrective action program as Callaway Action Request 201011233 and the licensee implemented procedure changes to address the potential for post-loss of coolant accident containment sump fluids being injected back to the refueling water storage tank.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of procedure quality and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding screened to a Phase 2 significance determination because it involved a potential loss of safety function. A Region IV senior reactor analyst performed a bounding Phase 3 significance determination and determined that the finding was of very low safety significance (Green). The very short exposure period coupled with the availability of equipment needed for other initiating events (other than small and medium loss of coolant accidents) helped to mitigate the significance. The dominant core damage sequences included small and medium break loss of coolant accidents, and the failure of emergency core cooling pumps in the recirculation mode. This finding was determined not to have a crosscutting aspect since the performance deficiency is not reflective of current performance.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 24, 2011
 
Identified By: NRC Item Type: NCV NonCited Violation Scaffolding Installation Inadequacy The inspectors identified a noncited violation of Technical Specification 5.4.1.a for failure to properly implement Procedure MDP-ZZ-S0001, "Scaffolding Installation and Evaluation," Revision 26, when scaffolding was erected near or in contact with equipment in safety-related structures. On February 8 and March 16, 2011, the inspectors identified two locations where scaffold poles and a scaffold pin were less than the procedure required 1 inch from the auxiliary building vent line, the Train B emergency diesel lube oil drain line, and also essential service water system piping in the Train B diesel room. This issue was entered into the licensees corrective action program as Callaway Action Request 201102091.
The deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding was associated with the Mitigating Systems Cornerstone. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the component of corrective action program because the licensee did not have a low threshold for identifying scaffold issues.
Inspection Report# : 2011002 (pdf)
Significance: SL-IV Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report The team identified a Severity Level IV, noncited violation of 10 CFR 50.71, Maintenance of records, making of reports, paragraph (e) which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the updated safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Specifically, the licensee incorporated numerous errors in the updated safety analysis report associated with the descriptions of the onsite electrical power systems. The licensee has entered this violation into their corrective action program as Condition Reports 201101335 and 201102064.
The inspectors determined that the failure to update the updated safety analysis report as required by 10 CFR 50.71(e),
Maintenance of records, making of reports was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function.
The inspectors used the NRC Enforcement Policy, dated September 30, 2010, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was associated with a traditional enforcement violation.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Design the Emergency Diesel Generator Ground Fault Protection Circuitry The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, when designing the bypass circuitry for the emergency diesel generator ground fault trip function, the licensee failed to ensure that the associated electrical components were adequately designed for the continuous duty they would have to withstand under bypassed trip conditions. This could result in an ignition source and subsequent fire in the area under these conditions. This finding was entered into the licensees corrective action program as Condition Report 201102064.
 
The team determined that the failure to analyze the suitability of the emergency diesel generator components when protection features were bypassed was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design of these components could have prevented continued operation of the emergency diesel generator under ground fault conditions with the trip signal bypassed. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Specifically, the licensee revised the associated procedures to include these components in the combustible material exclusion zone. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Residual Heat Removal Flow Alarm Setpoint The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of March 3, 2011, the Mode 6 residual heat removal system low flow alarm setpoint did not adequately account for flow measurement uncertainties, and consequently was non-conservative. The licensee has entered the violation into their corrective action program as Condition Report 201101750.
The team determined that the failure to adequately analyze the uncertainty in measurement of residual heat removal system flow, and the impact of this failure, was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design basis analysis, and plant instrumentation, did not ensure that, while operating in Mode 6, the control room operators would be alerted whenever the residual heat removal system flow through the reactor coolant system was below the required value of 1000 gallons per minute. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011006 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure in the Establishment of a Turbine Driven Auxiliary Feedwater Pump Postmaintenance Test Procedure The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving a failure to follow procedures in the development of Procedure OTS-FC-0006, "TDAFW Pump Post-Maintenance Test Run on Aux Steam." Specifically, the licensee failed to incorporate turbine-driven auxiliary feedwater pump vendor manual precautions, limitations, and technical information in Procedure OTS-FC-0006, which resulted in the axial unloading, rolling element ball skidding, and subsequent degradation to the turbine-driven auxiliary feedwater pump inner outboard thrust bearing. Following discovery during planned maintenance and as immediate corrective actions, the licensee
 
declared the turbine-driven auxiliary feedwater pump inoperable, entered the applicable Technical Specification Limiting Condition for Operation, replaced the oil and bearings, restored the pump to operability, and initiated Callaway Action Request 201101042 to perform a root cause analysis.
This finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and
-~----------eh-a-ra-cterizationofFfhalngs," and was determined to be of very low safety significance (Green) because there was not a design or qualification deficiency that resulted in a loss of operability or functionality, it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time, it did not represent an actual loss of risk significant equipment, and it did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to ensure procedural adherence in the establishment of the turbine-driven auxiliary feedwater pump postmaintenance test procedure.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure for Turbine-Driven Auxiliary Feedwater Pump Postmaintenance Testing The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving five examples of failure to follow 2 Enclosure Procedure OTS-FC-0006, "TDAFW Pump Post-Maintenance Test Run on Aux Steam." Specifically, operators failed to follow an existing total flow precaution in Procedure OTS-FC-0006 which resulted in the axial unloading, rolling element ball skidding, and subsequent degradation to the turbine-driven auxiliary feedwater pump inner outboard thrust bearing. Following initial condition discovery during planned maintenance and as immediate corrective actions, the licensee declared the turbine-driven auxiliary feedwater pump inoperable, entered the applicable Technical Specification Limiting Condition for Operation, replaced the oil and bearings, restored the pump to operability, and initiated Callaway Action Request 201101042 to perform a root cause analysis.
These findings were more than minor because they affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding was evaluated Using Manual Chapter 0609.04, "Phase 1 -Initial Screening and Characterization of Findings." These findings were determined to be of very low safety significance (Green) because there was not a design or qualification deficiency that resulted in a loss of operability or functionality, they did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time, they did not represent an actual loss of risk significant equipment, and they did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to ensure procedural adherence in the implementation of the turbine-driven auxiliary feedwater pump postmaintenance test procedure.
Inspection Report# : 2011007 (pdf)
 
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Past Operability Determination The inspectors identified a non cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," regarding the licensee's failure to follow the requirements of Procedure APA-ZZ-00500, Appendix 3, "Past Operability and Reportability Evaluations." Specifically, the inspectors identified that the past operability evaluation for the turbine-driven auxiliary feedwater pump used a nonconservative calculation of mission time that did not take into account all design and licensing basis functions when determining the mission time. The licensee entered this issue into their corrective action program as Callaway Action Request 201102431 and updated its mission time anal}"sis to account for the turbine-driven auxiliary feedwater pump's specified safety function to bring the plant to a safe shutdown condition.
This finding is greater than minor because if left uncorrected, it would have the potential to lead to a more Significant safety concern because systems that may be inoperable may not be recognized and that it impacted the Mitigating Systems Cornerstone attribute of human performance in that the failure to accurately understand the auxiliary feedwater system mission time affected the mitigating systems objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 -Initial Screening and Characterization of Findings," the finding is determined to have very low safety significance because it did not result in the loss of safety function of any technical specification required equipment. The cause of this finding is related to the problem identification and resolution cross-cutting component of corrective action program because licensee personnel failed to thoroughly evaluate conditions adverse to quality and perform adequate operability determinations.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Calculate and Implement Conservative Safety Related Equipment Oil Leakage Operability Criteria The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for a failure to adequately determine safety related equipment oil leakage acceptance criteria that was used in operator logs. Specifically, the 2008 licensee fluid leak management program calculations to determine the mission time assessments related to oil leak rates of safety related pumps and motors were nonconservative when added to Procedure ODP-ZZ-0016E, Appendix 1, "Equipment Operator General Inspection Guide." The licensee evaluated this issue in Callaway Action Request 201102431 and calculated new conservative oil leak rates for the affected equipment.
This finding is more than minor because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to adequately evaluate and determine an appropriate lube oil leak rate to maintain safety related equipment operability affects the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings,"
and determined to be of very low safety significance since the as-found condition of the safety related equipment reviewed back to August 15, 2007, found no oil leakage rates that would have caused a loss of system safety function. This
 
finding was not reflective of current licensee performance and therefore, has no cross-cutting aspect.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance Schedule to Protect Safety-Related Equipment from Undetected Degraded Conditions The inspectors identified a noncited violation of Technical Specification 5.4.1 for a failure to adequately establish and implement procedures required by Regulatory Guide 1.33, Appendix A, Section 9, "Procedures for Performing Maintenance." Specifically, the preventative maintenance schedule to perform periodic lube oil analysis established an 18 month frequency without adequate justification resulting in a failure to promptly detect bearing degradation in the turbine-driven auxiliary feedwater pump. The licensee evaluated this issue in Callaway Action Request 201101042 and has corrective actions to review the lube oil analysis frequency and reduce it to at least a 9 month frequency-,-. _______________ _
This finding is more than minor because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to adequately evaluate and determine an appropriate lube oil monitoring schedule resulted in the failure to promptly detect a degraded bearing in the turbine-driven auxiliary feedwater pump affecting the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings,"
and determined to be of very low safety significance since the as-found condition of the degraded bearing would not have caused a loss of system safety function.
The finding has a cross-cutting aspect in the area of human performance associated with the decision making component, in that, the licensee failed to use conservative assumptions in the decision to extend the turbine-driven auxiliary feedwater pump lube oil monitoring interval to 18 months.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Cause Results in Missed Safety-Related Pump Overhaul The inspectors identified a noncited violation of 10 CFR Part 50, Appendix 8, Criterion XVI, "Corrective Action," associated with the licensee's failure to promptly identify and correct a condition adverse to quality.
Specifically, the licensee reduced the scope of preventative maintenance for the turbine-driven auxiliary feedwater pump overhaul during Refueling Outage 16 without proper justification, resulting in the failure to perform required pump maintenance. This issue was entered into the licensee's corrective action program as Callaway Action Request 201102407 and the pump has been scheduled to be overhauled during the next refueling outage.
This finding is more than minor because, if left uncorrected, corrective action deficiencies would have the potential to lead to a more significant safety concern.
The failure to perform required maintenance could allow equipment degradation affecting the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined to be of very low
 
safety significance because the degraded condition did not result in a loss of operability or functionality. The inspectors determined that the finding has a cross-cutting aspect in the area of human performance associated with the work control component because the licensee does not appropriately coordinate work activities by incorporating actions to address the impact of changes to work scope on the plant such that nuclear safety is supported.
Inspection Report# : 2011007 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Operability Determination Procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to follow Procedure APA ZZ 00500, Appendix 1, Operability and Functionality Determinations. On the morning of September 23, 2010, Callaway engineering was informed that a concern existed that the safety related portion of the component cooling water system safety function could be affected by a guillotine break at the nonsafety/nonseismic boundary for supply and return piping to the radwaste building. The inspectors determined that the licensee staff did not engage the shift manager early enough and the shift manager did not adequately challenge the basis describing the nonconforming condition as acceptable. The shift manager allowed the component cooling water system to be in an indeterminate state of operability for over two hours without putting compensatory measures in place as described in Procedure APA ZZ 00500, Appendix 1. This issue was entered into the licensees corrective action program as Callaway Action Request 201010739.
This finding was determined to be greater than minor because it impacted the mitigating systems cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as requiring a Phase 3 analysis. The NRC senior risk analyst determined that because ?CDF was less than 1E-6 and ?LERF was not a significant contributor to risk, this finding was of very low safety significance, Green. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations.
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate, Untimely Corrective Actions for a Containment Spray System Condition Adverse to Quality The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
associated with the licensees failure to promptly identify and correct a boric acid leak on the containment spray system, a condition adverse to quality. During a plant walkdown on October 14, 2010, the inspectors noted the continued existence of a boric acid leak on the flow element above the discharge of the train A containment spray pump. Further inspection revealed the leak was first identified on February 16, 2009. The inspectors found that nearly twenty months after initial identification, the repair plan for the leak had not been assigned a scheduled date. The failure to promptly correct the leak was directly caused by a lack of coordination between the engineering and outage planning departments. This issue was entered into the licensees corrective action program as Callaway Action Request 201010263. Immediate corrective action included scheduling the repair for January 2011.
This finding is more than minor because, if left uncorrected, programmatic work control and corrective action deficiencies would have the potential to lead to a more significant safety concern. This finding affected the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded condition did not result in a loss of operability or functionality. The inspectors determined that the finding has a crosscutting aspect in the area of human performance because the licensee work practices did not ensure supervisory and management oversight of work activities, such that nuclear safety was supported.
 
Inspection Report# : 2010005 (pdf)
Significance:      Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Enter Condition Adverse to Quality Associated with Emergency Diesel Generator Jacket Water Keep Warm Pump into the Corrective Action Program The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to follow the requirements of Callaway Procedure APA ZZ 00500, Corrective Action Program, associated with a degraded train B emergency diesel generator jacket water keep warm pump. On November 6, 2010, the supply breaker to the train B emergency diesel generator jacket water keep warm pump tripped unexpectedly causing the engine to become inoperable. During follow-up investigation, the inspectors found that a March 31, 2009 motor circuit evaluation was performed that showed a step decrease in insulation resistance from 10,250 Mega-ohms to 3.5 Mega-ohms. The degradation was at a sufficient rate such that there was a reasonable doubt the motor would continue to be reliable until the next performance of the motor circuit evaluation. The licensee failed to recognize this degradation and, as a result, did not initiate a Callaway action request to evaluate the condition. This issue was entered into the licensees corrective action program as Callaway Action Request 201010654.
This finding is greater than minor because if left uncorrected, the failure to fully utilize the corrective action program could become a more significant safety concern. The inspectors determined that this finding impacted the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as having very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. The cause of this finding is related to the problem identification and resolution crosscutting component of the corrective action program because licensee personnel failed to implement a corrective action program with a low threshold for identifying issues.
Inspection Report# : 2010005 (pdf)
Significance:      Nov 05, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Degraded Conditions in Essential Service Water System in a Timely Manner The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to correct in a timely manner degraded conditions affecting the essential service water system. Specifically, the licensee failed to resolve the combined effects of corrosion and waterhammer events resulting in system leaks. The licensee has experienced the waterhammer events since initial plant startup and has been experiencing problems with corrosion since the mid 1990s. As corrective actions for this issue, the licensee plans to implement two system modifications next refueling outage to mitigate the impacts of waterhammer events. This noncited violation was entered into the corrective action program as Callaway Action Request 201010635.
The issue was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of human performance associated with the component of resources because the licensee did not maintain the plant to minimize long-standing equipment issues.
Inspection Report# : 2010006 (pdf)
 
Significance:        Nov 05, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Repetitive Failures in Steam Generator Atmospheric Dump Valves in a Timely Manner The team identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving the failure to promptly correct deficiencies affecting the steam generator atmospheric steam dump valves. In 2002, system engineers identified that the valves current-to-pressure transducers were experiencing degradation because they were subjected to high vibration, and a proposed modification to move the transducers to a low vibration area occurred in 2006. The licensee experienced several additional failures in 2009 and determined that the reliable life of the transducers was 18 months in the high vibration areas. As of the date of the inspection, only one transducer of the four had been moved to a low vibration location, and the team determined that corrective actions for this condition adverse to quality have not been timely. The licensee plans to implement modifications to relocate the remaining three transducers to a lower vibration environment in 2011. The issue was entered into the licensees corrective action program as Callaway Action Request 200910153.
This issue was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors found that even though the steam generator atmospheric steam dump valves were not able to meet their technical specification surveillance requirements of achieving the full open position the valves would open sufficiently to meet its intended safety function. Therefore, the issue was of very low safety significance since it was a design or qualification deficiency confirmed not to result in a loss of functionality. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to maintain long term plant safety by minimization of long-standing equipment issues associated with steam generator atmospheric steam dump valve current-to-pressure transducers.
Inspection Report# : 2010006 (pdf)
Significance:        Nov 05, 2010 Identified By: NRC Item Type: FIN Finding Failure to Follow the Corrective Action Program Procedure The team identified a finding involving the licensees failure to follow the corrective action program procedure for assigning significance levels to Callaway action requests. This deficiency resulted in the licensees failure to adequately evaluate the cause and extent of condition for a number of issues, and in some examples resulted in recurrences of the issues. In one example the licensee identified a jacket water leak on Emergency Diesel Generator B in 2008. This significant condition adverse to quality was assigned a Significance Level 3 which only required a lower tier cause evaluation, when the procedure identified a significant condition adverse to quality as an example of a Significance Level 1. The team identified additional examples involving degraded safety-related equipment and security-related issues. As corrective action, the licensee entered the issue into its corrective action program as Callaway Action Request 201010472.
This issue was determined to be greater than minor because if left uncorrected, the issue could become a more significant safety concern. The inspectors determined that the issue involving Callaway Action Request 200812985, the failure of emergency diesel generator train B due to a leak in the jacket water system, was of very low safety significance because it was bounded by the significance of NCV 05000483/2009007-01, Failure to Ensure Suitable Replacement Parts Essential for Emergency Diesel Generator Train B.
The team evaluated the issue involving Callaway Action Request 200810379, the failure of engineered safety feature power supply SA036E, using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. This issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events.
The team also evaluated several security-related examples of this finding that are described in Enclosure 2 of this
 
letter. These security issues were also determined to be of very low security significance. Based on the sensitivity of security issues, Enclosure 2 is not publicly available because it contains security-related information.
This finding has a crosscutting aspect in the area of human performance associated with the component of training because training was needed for the screening committee to better understand a significant condition adverse to quality and to better understand the significance of security issues.
Inspection Report# : 2010006 (pdf)
Significance: SL-IV Sep 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Accurately Report a Condition that Could Have Prevented Fulfillment of a Safety Function The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), "Licensee Event Report System," for failure to report simultaneous inoperability of two steam generator atmospheric steam dump valves as a condition that could have prevented fulfillment of a safety function. On February 8, 2010, AmerenUE submitted Licensee Event Report 05000483/2009-005-00 to document that steam generator atmospheric steam dump valve ABPV0002 was out of service longer than allowed by Technical Specification 3.7.4, "Atmospheric Steam Dump Valves (ASDs)." The licensee event report also documented a period where valve ABPV0002 inoperability overlapped the inoperability of steam generator atmospheric steam dump valve ABPV0003. Callaway Final Safety Analysis Report Section 15.6.3.2.2.p. stated that all three intact steam generator atmospheric steam dump valves are credited in the cool down for a steam generator tube rupture. The inspectors determined that the licensee failed to adequately evaluate the reportability of having simultaneous inoperability of two steam generator atmospheric steam dump valves as a safety system functional failure. This issue was entered into the licensees corrective action program as Callaway Action Request 201006086 and on September 29, 2010, the licensee submitted Licensee Event Report 05000483/2009-005-001 to correct the reporting error.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was strictly associated with a traditional enforcement violation.
Inspection Report# : 2010004 (pdf)
Significance:      Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Lack of Maintenance as Cause of Diesel Generator Failure The inspectors identified a Green noncited violation of 10 Part 50, Appendix B, Criterion V, for the failure to accomplish a root cause evaluation in accordance with station procedures. Specifically, the licensee failed to identify and document that implementing Fairbanks Morse Owners Group recommended maintenance would have had a high likelihood of preventing the March 30, 2010, emergency diesel generator failure. As a result, the licensee did not classify the addition of maintenance on the governor and the governor drive as a corrective action, and the lack of maintenance was not evaluated for extent of condition and corrective actions, as applicable. This issue has been entered into the licensees corrective action program as Callaway Action Request 201008405.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events. Specifically, the evaluation failed to discover the lack of maintenance on the diesel governor and drive and the licensee failed to classify the maintenance as necessary. In addition, there was a potential for other recommended maintenance not being performed on mitigating equipment due to not evaluating the extent of condition and cause. Using NRC Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance because the finding did not result in the loss of safety function for mitigating equipment. This finding has a crosscutting aspect in the problem identification and resolution area associated with the operating experience component, in that the
 
licensee failed to evaluate operating experience applicable to the root cause in a systematic and timely manner.
Inspection Report# : 2010007 (pdf)
Barrier Integrity Significance:      Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Establish Test Program for Isolation Valves in Post-LOCA Recirculation Flowpath The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for failure to adequately demonstrate that the seat leakage of centrifugal charging pump and safety injection pump suction isolation valves remained within acceptable limits. These valves have a combined allowable leakage rate of three gallons per minute to ensure that offsite thyroid and whole body doses remain within regulatory limits. Since the flowpaths have isolation valves for which seat leakage is limited to a specific maximum amount, the inspectors identified that they should be considered Category A valves as specified in ASME OM Code which requires the valves be tested at least once every two years. At the end of the inspection period, the licensee was planning a recurring surveillance test to verify seat leakage for these valves is within acceptable limits. This issue was entered into the licensees corrective action program as Callaway Action Request 201104577.
This finding was greater than minor because it was associated with the Barrier Integrity Cornerstone attribute of configuration control and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to represent an actual open pathway in the physical integrity of reactor containment. Using Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since the nontested flowpath would be comparable to small lines (less than 1 2 inches in diameter) and would not generally contribute to large early release frequency. This finding did not have a cross-cutting aspect since the error associated with the inservice testing program was not reflective of current licensee performance.
Inspection Report# : 2011003 (pdf)
Significance:      Sep 23, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Adequate Administration Controls for Failed Containment Isolation Valve The inspectors identified a green noncited violation of Technical Specification 3.6.3, "Containment Isolation Valves,"
after the licensee failed to implement adequate administrative controls following the failure of valve EGHV0059. On August 10, 2010, containment isolation valve EGHV0059 failed to indicate full closed in the control room. The licensee declared the valve inoperable and isolated the affected penetration flow path. To ensure reactor coolant pump cooling the licensee unisolated the penetration by opening valve EGHV0131 and placing it under administrative controls. The on-shift operations technician was assigned to isolate the penetration in the event containment isolation was required. The resident inspectors found the licensees administrative controls were not consistent with the requirements in the technical specification bases which required a dedicated operator at the valve. The licensee then stationed a dedicated operator at valve EGHV0131 while repairs were conducted on valve EGHV0059. This issue was entered into the licensees corrective action program as Callaway Action Request 201007644.
This finding is more than minor because it was associated with the Barrier Integrity Cornerstone attribute of procedural quality and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to represent an actual open pathway in the physical integrity of reactor containment. Using Manual Chapter 0609, Appendix H,
 
"Containment Integrity Significance Determination Process," the issue was determined to be a Type B finding of very low safety significance since the containment penetration was associated with a closed system and would generally not contribute to large early release frequency. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure procedures used for addressing administrative controls were accurate and consistent with the technical specification bases [H.2(c)].
Inspection Report# : 2010004 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:        May 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Recipients License Conditions Prior to shipping Special Nuclear Material The inspectors identified a noncited violation of 10 CFR 70.42 (c) for failure to verify that a recipient of special nuclear material was authorized to receive the quantity of material shipped. This finding was determined to be of very low safety significance. Specifically, On June 15, 2010, the licensee shipped laundry contaminated with radioactive material to a state licensed processing facility in Alabama. The licensee verified that the processing facility was licensed to handle the material being shipped, but failed to verify that the recipients license authorized the quantity of material shipped. The licensee notified the Alabama licensee and proposed a revision to the shipping procedures. This violation was entered into the licensees corrective action program as Callaway Action Request 201104385.
This finding was greater than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process (transportation program), and affected the cornerstone objective, in that, license conditions were violated and these conditions are in place, in part, to control exposure to radiation. Using the public radiation safety significance determination process, the inspectors determined the finding had very low safety significance because (1) radiation limits were not exceeded, (2) there was no breach of a package during transit, (3) it did not involve a certificate of compliance issue, (4) it was not a low level burial ground nonconformance, and (5) it did not involve a failure to make notifications or provide emergency information. This finding had a crosscutting aspect in the area of human performance, resources component, because licensee procedures were inadequate to ensure proper shipping of radioactive material and that license conditions were not violated.
Inspection Report# : 2011003 (pdf)
Significance: SL-IV May 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Periodically Update the Final Safety Analysis Report The inspectors identified a noncited violation of 10 CFR 50.71 Maintenance of Records, because the licensee failed to update their Final Safety Analysis Report with submittals that include the effects of a change made to the facility.
Specifically, the licensee built the old steam generator storage facility on the owner controlled area for long-term radwaste storage of four decommissioned steam generators and failed to update the Final Safety Analysis Report to include these changes to the facility. This issue was entered in the licensees corrective action program as Callaway Action Request 201104434.
This issue was dispositioned using traditional enforcement because it had the potential for impacting the NRCs
 
ability to perform its regulatory function. The finding is more than minor because it has a material impact on licensed activities in that the four decommissioned steam generators, with a significant radioactive source term, have been relocated from the plant radiological controlled area to the owner controlled area. In addition, the radwaste management program has been affected because the licensee determined that this low-level radwaste facility will store these large components until an appropriate facility for disposal can be determined. The finding is characterized as a Severity Level IV noncited violation in accordance with NRC Enforcement Policy, Section 6.1, and was treated as a noncited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy.
Inspection Report# : 2011003 (pdf)
Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Nov 05, 2010 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The team concluded that the corrective action program at the Callaway Plant was performing in a satisfactory manner to ensure safe plant operations. However, the team identified a number of instances in which the licensee did not follow its procedural guidance for assigning significance levels to problems identified and, as a result, did not adequately evaluate the causes and/or extent of conditions resulting in several repetitive issues.
The inspectors determined that the licensee evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. The inspectors noted that operating experience was considered in cause evaluations.
The team determined that the licensee had a healthy safety-conscious work environment in that workers felt free to raise safety concerns without fear of retaliation using all avenues available.
Inspection Report# : 2010006 (pdf)
Last modified : October 14, 2011
 
Callaway 3Q/2011 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Implement a Plant Safety System Test Procedure A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified when the licensees failure to correctly follow a test procedure resulted in a negative reactivity excursion due to excessive boration. On May 27, 2011, with the Callaway Plant at 100 percent power, maintenance was in progress to perform a functional test of the plants safety system trip actuating devices. During the test the instrument maintenance technicians failed to place the mode selector switch in the test position. This resulted in switching the charging pump suction from the volume control tank to the refueling water storage tank. The inadvertent actuation resulted in a reactivity excursion that required lowering main turbine power and reactor power to about 92 percent. The crew stabilized the plant and returned critical parameters to their normal control bands. The licensee entered this issue in the corrective action program as Callaway Action Request 201104451.
This finding is more than minor because it was associated with the configuration control attribute of the Initiating Events Cornerstone and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions will not be available. This finding had a cross-cutting aspect in the area of human performance associated with the work practices component because the instrument maintenance technicians failed to adequately use human error prevention techniques, such as self- and peer-checking to ensure that work activities are performed safely Inspection Report# : 2011003 (pdf)
Mitigating Systems Significance:        Sep 23, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Implement Plant Maintenance Procedures The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to correctly follow maintenance procedures which resulted in a failure of motor-operated valve EFHV0065 associated with the ultimate heat sink train A cooling tower. To perform its safety function the valve must be capable of being closed. On September 15, 2010, the mechanical maintenance department removed and rebuilt the actuator for the motor-operated valve. The valve actuator stop nuts were not set correctly and remained set outside the range of the electrical limits due to electrical maintenance workers failing to complete the procedure and work instructions initiated by the mechanical department. On June 22, 2011, an attempt to manually align essential service water return over the train A safety-related cooling tower failed when the motor-operated valve was manually positioned past the zero percent open position due to the improperly set stop nuts. This disengaged the valve operator worm from its worm gear, opened the valve, and rendered the valve being incapable of being closed. The immediate corrective action to replace the valve actuator was completed on June 23, 2011. The licensee initiated Callaway Action Request 201105074 to evaluate cause and extent-of-condition and specify corrective actions.
This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating
 
Systems Cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work controls component because the mechanical and electrical maintenance technicians failed to adequately maintain interfaces to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2011004 (pdf)
Significance:      Sep 23, 2011 Identified By: Self-Revealing Item Type: FIN Finding Failure to Evaluate Breaker Relay Settings Results in Partial Loss of Station Blackout Response Capability The inspectors reviewed a self-revealing finding for the failure of AmerenUE engineering personnel to correctly establish the relay settings for the alternate emergency power supply diesel output breakers. On August 21, 2011, Callaway Plant experienced a loss of power to the alternate emergency power supply diesel bus PA05. This resulted in all four alternate emergency power supply diesels starting; however, the number three diesel output breaker immediately tripped open. The licensee determined that the breakers protective relaying was improperly set. Further investigation by AmerenUE discovered that all four of the diesel output breakers had incorrect settings. The incorrect settings occurred due to the limited range of the relay chosen for the application and the engineering recommendations that prioritized protecting the diesel over limiting the margin to unintended breaker trips. Callaway engineering reviews had not identified the low margin to unintended trips. The licensee initiated corrective actions associated with Callaway Action Request 201106701 to change the differential current relay settings.
This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the finding is of very low safety significance because it was a design deficiency that did not result in a loss of system safety function, did not represent an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk-significant per 10 CFR 50.65, for greater than 24 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of problem identification and resolution because the licensee failed to implement a corrective action program with a low threshold for identifying issues commensurate with their safety significance.
Inspection Report# : 2011004 (pdf)
Significance:      Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Flooding Analysis for Room 3101 The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to provide adequate design control measures for verifying the adequacy of the flooding analysis associated with the 2009 modification that replaced essential service water carbon steel piping with high density polyethylene piping. The licensee did not update the flooding analysis of record to consider potential failures in the new piping. The licensee generated Callaway Action Request 201102957 to develop a means to evaluate the relative stresses associated with the new pipe.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding required a Phase 2 significance determination. Using the presolved worksheet from the Risk Informed Inspection Notebook for the Callaway Station, Revision 2.01a, the finding was red, which warranted further review. Therefore, a senior reactor analyst performed a bounding Phase 3 significance determination. The bounding change to the core damage frequency was
 
approximately 4.1E-7 (Green). This was impacted significantly by the very small amount of new piping in the room.
This finding was determined to have a cross-cutting aspect in the area of Problem Identification and Resolution associated with the corrective action component in that the licensee did not thoroughly evaluate the extent of condition when the residents challenged the flooding calculation in December 2010 such that the resolutions addressed causes and extent of conditions, as necessary Inspection Report# : 2011003 (pdf)
Significance:      Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Analyze Refueling Water Storage Tank Level Transmitters for High-Energy Line Break The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to adequately evaluate a potential high-energy line break in nonseismically qualified auxiliary steam piping in the refueling water storage tank valve house. The harsh environment from a high-energy line break had the potential to impact safety related level transmitters associated with the refueling water storage tank. Following identification of this issue by the inspectors, the licensee analyzed the nonnuclear auxiliary piping to ensure it could withstand safe shutdown earthquake loadings which allowed high-energy line breaks at intermediate locations to be excluded. This issue was entered into the licensees corrective action program as Callaway Action Request 201102588.
This finding is greater than minor because it is associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, this finding is determined to be of very low safety significance since subsequent evaluation concluded the issue was a design or qualification deficiency confirmed not to result in loss of operability or functionality. This finding did not have a cross-cutting aspect since the error associated with the high-energy line break analysis was not reflective of current licensee performance.
Inspection Report# : 2011003 (pdf)
Significance:      Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Document Reasonable Expectation of Operability for Equipment Supported b y the Class 1E Air Conditioning Units The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to adequately evaluate past operability associated with the Class 1E electrical equipment air conditioning unit. The inspectors identified that Revision 1 and 2 to Callaway Action Request 200800615 incorrectly concluded that the equipment supported by the Class 1E electrical equipment air conditioning unit train B was operable with the units cooling water flow control valve in manual. This issue was entered into the licensees corrective action program as Callaway Action Request 201102565.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened to a Phase 2 significance determination because it involved the loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a bounding Phase 3 significance determination and determined that the finding was of very low safety significance (Green). The very short exposure period coupled with the availability of train A equipment helped to mitigate the significance. The dominant core damage sequences included a loss of main feedwater initiating event; the loss of train B electrical power; and various failures of auxiliary feedwater. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions including verifying the validity of the underlying assumptions when performing operability/reportability evaluations.
Inspection Report# : 2011002 (pdf)
 
Significance: SL-IV Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report Inoperability of Class IE Electrical Equipment for a Period Greater than Allowed by the Plant's Technical Specifications The inspectors identified a IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for failure to report inoperability of Class 1E electrical equipment for a period greater than allowed by the plants technical specifications. The licensee determined there were no prior instances where the Class 1E electrical equipment air conditioning units were inoperable greater than the technical specification allowed completion time of the supported equipment. The inspectors reviewed the licensees reportability evaluation and identified that the event described in Callaway Action Request 200800615 resulted in a period where the Class 1E electrical equipment air conditioning unit train B was inoperable for approximately 37 hours which exceeded the technical specification allowed completion time of the equipment supported by the Class 1E electrical equipment and constituted a condition which was prohibited by the plant's technical specifications and should have been reported in a licensee event report.
This issue was entered into the licensees corrective action program as Callaway Action Request 201011132.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because in order to perform its regulatory function, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was strictly associated with a traditional enforcement violation.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Containment Spray Test Procedure Potentially Creates an Unanalyzed Condition The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to provide adequate procedural guidance for testing of containment spray pumps. The inspectors reviewed a licensee evaluation of the acceptability of their existing containment spray pump testing procedure and found that it failed to adequately address the underlying technical issues because it relied on operators recognizing the diversion flow path and focused on the operability of the containment spray system and not the ability to maintain the long term cooling function of the emergency core cooling system. Additionally, the inspectors identified that the procedure would have provided a diversion flow path of post-accident sump fluids back to the refueling water storage tank exceeding those currently analyzed in the Callaway licensing bases. This issue was entered into the licensees corrective action program as Callaway Action Request 201011233 and the licensee implemented procedure changes to address the potential for post-loss of coolant accident containment sump fluids being injected back to the refueling water storage tank.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of procedure quality and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding screened to a Phase 2 significance determination because it involved a potential loss of safety function. A Region IV senior reactor analyst performed a bounding Phase 3 significance determination and determined that the finding was of very low safety significance (Green). The very short exposure period coupled with the availability of equipment needed for other initiating events (other than small and medium loss of coolant accidents) helped to mitigate the significance. The dominant core damage sequences included small and medium break loss of coolant accidents, and the failure of emergency core cooling pumps in the recirculation mode. This finding was determined not to have a crosscutting aspect since the performance deficiency is not reflective of current performance.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation
 
Scaffolding Installation Inadequacy The inspectors identified a noncited violation of Technical Specification 5.4.1.a for failure to properly implement Procedure MDP-ZZ-S0001, "Scaffolding Installation and Evaluation," Revision 26, when scaffolding was erected near or in contact with equipment in safety-related structures. On February 8 and March 16, 2011, the inspectors identified two locations where scaffold poles and a scaffold pin were less than the procedure required 1 inch from the auxiliary building vent line, the Train B emergency diesel lube oil drain line, and also essential service water system piping in the Train B diesel room. This issue was entered into the licensees corrective action program as Callaway Action Request 201102091.
The deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding was associated with the Mitigating Systems Cornerstone. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the component of corrective action program because the licensee did not have a low threshold for identifying scaffold issues.
Inspection Report# : 2011002 (pdf)
Significance: SL-IV Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report The team identified a Severity Level IV, noncited violation of 10 CFR 50.71, Maintenance of records, making of reports, paragraph (e) which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the updated safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Specifically, the licensee incorporated numerous errors in the updated safety analysis report associated with the descriptions of the onsite electrical power systems. The licensee has entered this violation into their corrective action program as Condition Reports 201101335 and 201102064.
The inspectors determined that the failure to update the updated safety analysis report as required by 10 CFR 50.71(e),
Maintenance of records, making of reports was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function.
The inspectors used the NRC Enforcement Policy, dated September 30, 2010, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was associated with a traditional enforcement violation.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Design the Emergency Diesel Generator Ground Fault Protection Circuitry The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, when designing the bypass circuitry for the emergency diesel generator ground fault trip function, the licensee failed to ensure that the associated electrical components were adequately designed for the continuous duty they would have to withstand under bypassed trip conditions. This could result in an ignition source and subsequent fire in the area under these conditions. This finding was entered into the licensees corrective action program as Condition Report 201102064.
The team determined that the failure to analyze the suitability of the emergency diesel generator components when
 
protection features were bypassed was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design of these components could have prevented continued operation of the emergency diesel generator under ground fault conditions with the trip signal bypassed. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Specifically, the licensee revised the associated procedures to include these components in the combustible material exclusion zone. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Residual Heat Removal Flow Alarm Setpoint The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of March 3, 2011, the Mode 6 residual heat removal system low flow alarm setpoint did not adequately account for flow measurement uncertainties, and consequently was non-conservative. The licensee has entered the violation into their corrective action program as Condition Report 201101750.
The team determined that the failure to adequately analyze the uncertainty in measurement of residual heat removal system flow, and the impact of this failure, was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design basis analysis, and plant instrumentation, did not ensure that, while operating in Mode 6, the control room operators would be alerted whenever the residual heat removal system flow through the reactor coolant system was below the required value of 1000 gallons per minute. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011006 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure in the Establishment of a Turbine Driven Auxiliary Feedwater Pump Postmaintenance Test Procedure The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving a failure to follow procedures in the development of Procedure OTS-FC-0006, "TDAFW Pump Post-Maintenance Test Run on Aux Steam." Specifically, the licensee failed to incorporate turbine-driven auxiliary feedwater pump vendor manual precautions, limitations, and technical information in Procedure OTS-FC-0006, which resulted in the axial unloading, rolling element ball skidding, and subsequent degradation to the turbine-driven auxiliary feedwater pump inner outboard thrust bearing. Following discovery during planned maintenance and as immediate corrective actions, the licensee declared the turbine-driven auxiliary feedwater pump inoperable, entered the applicable Technical Specification Limiting Condition for Operation, replaced the
 
oil and bearings, restored the pump to operability, and initiated Callaway Action Request 201101042 to perform a root cause analysis.
This finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and
-~----------eh-a-ra-cterizationofFfhalngs," and was determined to be of very low safety significance (Green) because there was not a design or qualification deficiency that resulted in a loss of operability or functionality, it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time, it did not represent an actual loss of risk significant equipment, and it did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to ensure procedural adherence in the establishment of the turbine-driven auxiliary feedwater pump postmaintenance test procedure.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure for Turbine-Driven Auxiliary Feedwater Pump Postmaintenance Testing The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving five examples of failure to follow 2 Enclosure Procedure OTS-FC-0006, "TDAFW Pump Post-Maintenance Test Run on Aux Steam." Specifically, operators failed to follow an existing total flow precaution in Procedure OTS-FC-0006 which resulted in the axial unloading, rolling element ball skidding, and subsequent degradation to the turbine-driven auxiliary feedwater pump inner outboard thrust bearing. Following initial condition discovery during planned maintenance and as immediate corrective actions, the licensee declared the turbine-driven auxiliary feedwater pump inoperable, entered the applicable Technical Specification Limiting Condition for Operation, replaced the oil and bearings, restored the pump to operability, and initiated Callaway Action Request 201101042 to perform a root cause analysis.
These findings were more than minor because they affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding was evaluated Using Manual Chapter 0609.04, "Phase 1 -Initial Screening and Characterization of Findings." These findings were determined to be of very low safety significance (Green) because there was not a design or qualification deficiency that resulted in a loss of operability or functionality, they did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time, they did not represent an actual loss of risk significant equipment, and they did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to ensure procedural adherence in the implementation of the turbine-driven auxiliary feedwater pump postmaintenance test procedure.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011
 
Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Past Operability Determination The inspectors identified a non cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," regarding the licensee's failure to follow the requirements of Procedure APA-ZZ-00500, Appendix 3, "Past Operability and Reportability Evaluations." Specifically, the inspectors identified that the past operability evaluation for the turbine-driven auxiliary feedwater pump used a nonconservative calculation of mission time that did not take into account all design and licensing basis functions when determining the mission time. The licensee entered this issue into their corrective action program as Callaway Action Request 201102431 and updated its mission time anal}"sis to account for the turbine-driven auxiliary feedwater pump's specified safety function to bring the plant to a safe shutdown condition.
This finding is greater than minor because if left uncorrected, it would have the potential to lead to a more Significant safety concern because systems that may be inoperable may not be recognized and that it impacted the Mitigating Systems Cornerstone attribute of human performance in that the failure to accurately understand the auxiliary feedwater system mission time affected the mitigating systems objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 -Initial Screening and Characterization of Findings," the finding is determined to have very low safety significance because it did not result in the loss of safety function of any technical specification required equipment. The cause of this finding is related to the problem identification and resolution cross-cutting component of corrective action program because licensee personnel failed to thoroughly evaluate conditions adverse to quality and perform adequate operability determinations.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Calculate and Implement Conservative Safety Related Equipment Oil Leakage Operability Criteria The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for a failure to adequately determine safety related equipment oil leakage acceptance criteria that was used in operator logs. Specifically, the 2008 licensee fluid leak management program calculations to determine the mission time assessments related to oil leak rates of safety related pumps and motors were nonconservative when added to Procedure ODP-ZZ-0016E, Appendix 1, "Equipment Operator General Inspection Guide." The licensee evaluated this issue in Callaway Action Request 201102431 and calculated new conservative oil leak rates for the affected equipment.
This finding is more than minor because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to adequately evaluate and determine an appropriate lube oil leak rate to maintain safety related equipment operability affects the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings,"
and determined to be of very low safety significance since the as-found condition of the safety related equipment reviewed back to August 15, 2007, found no oil leakage rates that would have caused a loss of system safety function. This finding was not reflective of current licensee performance and therefore, has no cross-cutting aspect.
 
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance Schedule to Protect Safety-Related Equipment from Undetected Degraded Conditions The inspectors identified a noncited violation of Technical Specification 5.4.1 for a failure to adequately establish and implement procedures required by Regulatory Guide 1.33, Appendix A, Section 9, "Procedures for Performing Maintenance." Specifically, the preventative maintenance schedule to perform periodic lube oil analysis established an 18 month frequency without adequate justification resulting in a failure to promptly detect bearing degradation in the turbine-driven auxiliary feedwater pump. The licensee evaluated this issue in Callaway Action Request 201101042 and has corrective actions to review the lube oil analysis frequency and reduce it to at least a 9 month frequency-,-. _______________ _
This finding is more than minor because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to adequately evaluate and determine an appropriate lube oil monitoring schedule resulted in the failure to promptly detect a degraded bearing in the turbine-driven auxiliary feedwater pump affecting the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings,"
and determined to be of very low safety significance since the as-found condition of the degraded bearing would not have caused a loss of system safety function.
The finding has a cross-cutting aspect in the area of human performance associated with the decision making component, in that, the licensee failed to use conservative assumptions in the decision to extend the turbine-driven auxiliary feedwater pump lube oil monitoring interval to 18 months.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Cause Results in Missed Safety-Related Pump Overhaul The inspectors identified a noncited violation of 10 CFR Part 50, Appendix 8, Criterion XVI, "Corrective Action," associated with the licensee's failure to promptly identify and correct a condition adverse to quality.
Specifically, the licensee reduced the scope of preventative maintenance for the turbine-driven auxiliary feedwater pump overhaul during Refueling Outage 16 without proper justification, resulting in the failure to perform required pump maintenance. This issue was entered into the licensee's corrective action program as Callaway Action Request 201102407 and the pump has been scheduled to be overhauled during the next refueling outage.
This finding is more than minor because, if left uncorrected, corrective action deficiencies would have the potential to lead to a more significant safety concern.
The failure to perform required maintenance could allow equipment degradation affecting the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined to be of very low safety significance because the degraded condition did not result in a loss of operability or functionality. The inspectors determined that the finding has a
 
cross-cutting aspect in the area of human performance associated with the work control component because the licensee does not appropriately coordinate work activities by incorporating actions to address the impact of changes to work scope on the plant such that nuclear safety is supported.
Inspection Report# : 2011007 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Operability Determination Procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to follow Procedure APA ZZ 00500, Appendix 1, Operability and Functionality Determinations. On the morning of September 23, 2010, Callaway engineering was informed that a concern existed that the safety related portion of the component cooling water system safety function could be affected by a guillotine break at the nonsafety/nonseismic boundary for supply and return piping to the radwaste building. The inspectors determined that the licensee staff did not engage the shift manager early enough and the shift manager did not adequately challenge the basis describing the nonconforming condition as acceptable. The shift manager allowed the component cooling water system to be in an indeterminate state of operability for over two hours without putting compensatory measures in place as described in Procedure APA ZZ 00500, Appendix 1. This issue was entered into the licensees corrective action program as Callaway Action Request 201010739.
This finding was determined to be greater than minor because it impacted the mitigating systems cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this issue screened as requiring a Phase 3 analysis. The NRC senior risk analyst determined that because ?CDF was less than 1E-6 and ?LERF was not a significant contributor to risk, this finding was of very low safety significance, Green. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions when performing operability evaluations.
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate, Untimely Corrective Actions for a Containment Spray System Condition Adverse to Quality The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
associated with the licensees failure to promptly identify and correct a boric acid leak on the containment spray system, a condition adverse to quality. During a plant walkdown on October 14, 2010, the inspectors noted the continued existence of a boric acid leak on the flow element above the discharge of the train A containment spray pump. Further inspection revealed the leak was first identified on February 16, 2009. The inspectors found that nearly twenty months after initial identification, the repair plan for the leak had not been assigned a scheduled date. The failure to promptly correct the leak was directly caused by a lack of coordination between the engineering and outage planning departments. This issue was entered into the licensees corrective action program as Callaway Action Request 201010263. Immediate corrective action included scheduling the repair for January 2011.
This finding is more than minor because, if left uncorrected, programmatic work control and corrective action deficiencies would have the potential to lead to a more significant safety concern. This finding affected the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded condition did not result in a loss of operability or functionality. The inspectors determined that the finding has a crosscutting aspect in the area of human performance because the licensee work practices did not ensure supervisory and management oversight of work activities, such that nuclear safety was supported.
Inspection Report# : 2010005 (pdf)
 
Significance:      Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Enter Condition Adverse to Quality Associated with Emergency Diesel Generator Jacket Water Keep Warm Pump into the Corrective Action Program The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to follow the requirements of Callaway Procedure APA ZZ 00500, Corrective Action Program, associated with a degraded train B emergency diesel generator jacket water keep warm pump. On November 6, 2010, the supply breaker to the train B emergency diesel generator jacket water keep warm pump tripped unexpectedly causing the engine to become inoperable. During follow-up investigation, the inspectors found that a March 31, 2009 motor circuit evaluation was performed that showed a step decrease in insulation resistance from 10,250 Mega-ohms to 3.5 Mega-ohms. The degradation was at a sufficient rate such that there was a reasonable doubt the motor would continue to be reliable until the next performance of the motor circuit evaluation. The licensee failed to recognize this degradation and, as a result, did not initiate a Callaway action request to evaluate the condition. This issue was entered into the licensees corrective action program as Callaway Action Request 201010654.
This finding is greater than minor because if left uncorrected, the failure to fully utilize the corrective action program could become a more significant safety concern. The inspectors determined that this finding impacted the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue screened as having very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. The cause of this finding is related to the problem identification and resolution crosscutting component of the corrective action program because licensee personnel failed to implement a corrective action program with a low threshold for identifying issues.
Inspection Report# : 2010005 (pdf)
Significance:      Nov 05, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Degraded Conditions in Essential Service Water System in a Timely Manner The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to correct in a timely manner degraded conditions affecting the essential service water system. Specifically, the licensee failed to resolve the combined effects of corrosion and waterhammer events resulting in system leaks. The licensee has experienced the waterhammer events since initial plant startup and has been experiencing problems with corrosion since the mid 1990s. As corrective actions for this issue, the licensee plans to implement two system modifications next refueling outage to mitigate the impacts of waterhammer events. This noncited violation was entered into the corrective action program as Callaway Action Request 201010635.
The issue was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of human performance associated with the component of resources because the licensee did not maintain the plant to minimize long-standing equipment issues.
Inspection Report# : 2010006 (pdf)
Significance:      Nov 05, 2010
 
Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Repetitive Failures in Steam Generator Atmospheric Dump Valves in a Timely Manner The team identified a green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving the failure to promptly correct deficiencies affecting the steam generator atmospheric steam dump valves. In 2002, system engineers identified that the valves current-to-pressure transducers were experiencing degradation because they were subjected to high vibration, and a proposed modification to move the transducers to a low vibration area occurred in 2006. The licensee experienced several additional failures in 2009 and determined that the reliable life of the transducers was 18 months in the high vibration areas. As of the date of the inspection, only one transducer of the four had been moved to a low vibration location, and the team determined that corrective actions for this condition adverse to quality have not been timely. The licensee plans to implement modifications to relocate the remaining three transducers to a lower vibration environment in 2011. The issue was entered into the licensees corrective action program as Callaway Action Request 200910153.
This issue was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors found that even though the steam generator atmospheric steam dump valves were not able to meet their technical specification surveillance requirements of achieving the full open position the valves would open sufficiently to meet its intended safety function. Therefore, the issue was of very low safety significance since it was a design or qualification deficiency confirmed not to result in a loss of functionality. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to maintain long term plant safety by minimization of long-standing equipment issues associated with steam generator atmospheric steam dump valve current-to-pressure transducers.
Inspection Report# : 2010006 (pdf)
Significance:        Nov 05, 2010 Identified By: NRC Item Type: FIN Finding Failure to Follow the Corrective Action Program Procedure The team identified a finding involving the licensees failure to follow the corrective action program procedure for assigning significance levels to Callaway action requests. This deficiency resulted in the licensees failure to adequately evaluate the cause and extent of condition for a number of issues, and in some examples resulted in recurrences of the issues. In one example the licensee identified a jacket water leak on Emergency Diesel Generator B in 2008. This significant condition adverse to quality was assigned a Significance Level 3 which only required a lower tier cause evaluation, when the procedure identified a significant condition adverse to quality as an example of a Significance Level 1. The team identified additional examples involving degraded safety-related equipment and security-related issues. As corrective action, the licensee entered the issue into its corrective action program as Callaway Action Request 201010472.
This issue was determined to be greater than minor because if left uncorrected, the issue could become a more significant safety concern. The inspectors determined that the issue involving Callaway Action Request 200812985, the failure of emergency diesel generator train B due to a leak in the jacket water system, was of very low safety significance because it was bounded by the significance of NCV 05000483/2009007-01, Failure to Ensure Suitable Replacement Parts Essential for Emergency Diesel Generator Train B.
The team evaluated the issue involving Callaway Action Request 200810379, the failure of engineered safety feature power supply SA036E, using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. This issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events.
The team also evaluated several security-related examples of this finding that are described in Enclosure 2 of this letter. These security issues were also determined to be of very low security significance. Based on the sensitivity of security issues, Enclosure 2 is not publicly available because it contains security-related information.
 
This finding has a crosscutting aspect in the area of human performance associated with the component of training because training was needed for the screening committee to better understand a significant condition adverse to quality and to better understand the significance of security issues.
Inspection Report# : 2010006 (pdf)
Barrier Integrity Significance:      Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Establish Test Program for Isolation Valves in Post-LOCA Recirculation Flowpath The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for failure to adequately demonstrate that the seat leakage of centrifugal charging pump and safety injection pump suction isolation valves remained within acceptable limits. These valves have a combined allowable leakage rate of three gallons per minute to ensure that offsite thyroid and whole body doses remain within regulatory limits. Since the flowpaths have isolation valves for which seat leakage is limited to a specific maximum amount, the inspectors identified that they should be considered Category A valves as specified in ASME OM Code which requires the valves be tested at least once every two years. At the end of the inspection period, the licensee was planning a recurring surveillance test to verify seat leakage for these valves is within acceptable limits. This issue was entered into the licensees corrective action program as Callaway Action Request 201104577.
This finding was greater than minor because it was associated with the Barrier Integrity Cornerstone attribute of configuration control and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to represent an actual open pathway in the physical integrity of reactor containment. Using Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since the nontested flowpath would be comparable to small lines (less than 1 2 inches in diameter) and would not generally contribute to large early release frequency. This finding did not have a cross-cutting aspect since the error associated with the inservice testing program was not reflective of current licensee performance.
Inspection Report# : 2011003 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:      May 27, 2011 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Verify Recipients License Conditions Prior to shipping Special Nuclear Material The inspectors identified a noncited violation of 10 CFR 70.42 (c) for failure to verify that a recipient of special nuclear material was authorized to receive the quantity of material shipped. This finding was determined to be of very low safety significance. Specifically, On June 15, 2010, the licensee shipped laundry contaminated with radioactive material to a state licensed processing facility in Alabama. The licensee verified that the processing facility was licensed to handle the material being shipped, but failed to verify that the recipients license authorized the quantity of material shipped. The licensee notified the Alabama licensee and proposed a revision to the shipping procedures. This violation was entered into the licensees corrective action program as Callaway Action Request 201104385.
This finding was greater than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process (transportation program), and affected the cornerstone objective, in that, license conditions were violated and these conditions are in place, in part, to control exposure to radiation. Using the public radiation safety significance determination process, the inspectors determined the finding had very low safety significance because (1) radiation limits were not exceeded, (2) there was no breach of a package during transit, (3) it did not involve a certificate of compliance issue, (4) it was not a low level burial ground nonconformance, and (5) it did not involve a failure to make notifications or provide emergency information. This finding had a crosscutting aspect in the area of human performance, resources component, because licensee procedures were inadequate to ensure proper shipping of radioactive material and that license conditions were not violated.
Inspection Report# : 2011003 (pdf)
Significance: SL-IV May 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Periodically Update the Final Safety Analysis Report The inspectors identified a noncited violation of 10 CFR 50.71 Maintenance of Records, because the licensee failed to update their Final Safety Analysis Report with submittals that include the effects of a change made to the facility.
Specifically, the licensee built the old steam generator storage facility on the owner controlled area for long-term radwaste storage of four decommissioned steam generators and failed to update the Final Safety Analysis Report to include these changes to the facility. This issue was entered in the licensees corrective action program as Callaway Action Request 201104434.
This issue was dispositioned using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding is more than minor because it has a material impact on licensed activities in that the four decommissioned steam generators, with a significant radioactive source term, have been relocated from the plant radiological controlled area to the owner controlled area. In addition, the radwaste management program has been affected because the licensee determined that this low-level radwaste facility will store these large components until an appropriate facility for disposal can be determined. The finding is characterized as a Severity Level IV noncited violation in accordance with NRC Enforcement Policy, Section 6.1, and was treated as a noncited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy.
Inspection Report# : 2011003 (pdf)
Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance: N/A Nov 05, 2010 Identified By: NRC
 
Item Type: FIN Finding Problem Identification and Resolution The team concluded that the corrective action program at the Callaway Plant was performing in a satisfactory manner to ensure safe plant operations. However, the team identified a number of instances in which the licensee did not follow its procedural guidance for assigning significance levels to problems identified and, as a result, did not adequately evaluate the causes and/or extent of conditions resulting in several repetitive issues.
The inspectors determined that the licensee evaluated industry operating experience for relevance to the facility and entered applicable items in the corrective action program. The inspectors noted that operating experience was considered in cause evaluations.
The team determined that the licensee had a healthy safety-conscious work environment in that workers felt free to raise safety concerns without fear of retaliation using all avenues available.
Inspection Report# : 2010006 (pdf)
Last modified : January 04, 2012
 
Callaway 4Q/2011 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Separation of Stainless Steel and Carbon Steel Hand Files and Wire Brushes The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, for the failure to have procedures that ensured that hand files and wire brushes designated for stainless steel weld preparation were stored separately from hand files and wire brushes used on carbon steel. The licensee took corrective actions to remove the stainless steel designations from stainless steel tools that were mixed with tools used on carbon steel, established segregated locations in tool rooms for the separation of abrasive tools, and trained tool room attendants to properly store and mark abrasive tools designated for use on stainless steel. This issue was entered into the licensees corrective action program as Callaway Action Request 201108921.
Inspectors determined that the failure to assure that hand files and wire brushes designated for exclusive use on stainless steel were stored separately from tools used on other materials was a performance deficiency. This finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and, if left uncorrected, could become a more significant safety concern. Inspectors performed a Phase 1 screening in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance because the issue would not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. This finding has a cross-cutting aspect in the area of problem identification and resolution, associated with the corrective action program, because the licensee did not thoroughly evaluate problems such that the resolutions addressed causes and extent of conditions, as necessary. Specifically, the licensees response to Callaway Action Request 201107806 identified contaminated tools as the cause of rusting on the motor-driven auxiliary feed pump room cooler stainless steel piping, but the licensee took no further action to identify the cause of the contamination.
Inspection Report# : 2011005 (pdf)
Significance:        Jun 23, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Implement a Plant Safety System Test Procedure A self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, was identified when the licensees failure to correctly follow a test procedure resulted in a negative reactivity excursion due to excessive boration. On May 27, 2011, with the Callaway Plant at 100 percent power, maintenance was in progress to perform a functional test of the plants safety system trip actuating devices. During the test the instrument maintenance technicians failed to place the mode selector switch in the test position. This resulted in switching the charging pump suction from the volume control tank to the refueling water storage tank. The inadvertent actuation resulted in a reactivity excursion that required lowering main turbine power and reactor power to about 92 percent. The crew stabilized the plant and returned critical parameters to their normal control bands. The licensee entered this issue in the corrective action program as Callaway Action Request 201104451.
This finding is more than minor because it was associated with the configuration control attribute of the Initiating Events Cornerstone and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety
 
significance since it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions will not be available. This finding had a cross-cutting aspect in the area of human performance associated with the work practices component because the instrument maintenance technicians failed to adequately use human error prevention techniques, such as self- and peer-checking to ensure that work activities are performed safely Inspection Report# : 2011003 (pdf)
Mitigating Systems Significance:      Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Simulator Fidelity The inspectors identified a non-cited violation of 10 CFR Part 55.46(c), Plant-Referenced Simulators, for failure of the licensee to ensure that the plant-referenced simulator demonstrated expected plant response to transient and accident conditions to which the simulator has been designed to respond. Specifically, the licensee failed to ensure simulator modeling of power-operated relief valve and pressurizer safety valve operation was consistent with the actual plant, introducing the potential for negative operator training. Due to errors made in modeling updates after steam generator replacement in 2005, each pressurizer safety valve was sized in the simulator to allow approximately 3.3 times higher than the design flow rate in the actual plant, and each power operated relief valve was sized to allow approximately 3.5 times higher than the design flow rate capacity provided in the actual plant. The licensee documented their corrective actions for this issue in Callaway Action Request 201101255.
The failure of the licensees simulator staff to ensure that the plant-referenced simulator demonstrated expected plant response to transient and accident conditions for which the simulator has been designed to respond was a performance deficiency. The performance deficiency is more than minor because it adversely impacted the human performance attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the performance deficiency could have become more significant in that training on related accident scenarios could have a negative impact on how licensed operators would respond to an actual event in the control room. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process, the finding was determined to have very low safety significance (Green) because there was no actual event at the plant similar to the simulator scenario where inappropriate actions were taken in the control room based on training with incorrectly sized components in the simulator. This finding has no cross-cutting aspect assigned because the cause was not representative of current licensee performance.
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: NRC Item Type: FIN Finding Failure to Conduct Simulator Testing in Accordance with ANSI/ANS 3.5-1998 The inspectors identified a finding associated with the conduct of simulator performance testing because the licensee was not testing in accordance with the standards of ANSI/ANS 3.5-1998. Specifically, the licensee did not include relief valve flow in their 2010 test of transient (10) of ANSI/ANS 3.5-1998, Appendix B, Section B3.2.1, "Slow Primary System Depressurization to Saturated Condition with Pressurizer Relief or Safety Valve Stuck Open. The licensee initiated corrective action documented in Callaway Action Request 201107912.
Conducting simulator performance testing that was not in accordance with the ANSI/ANS 3.5-1998 standard was a performance deficiency. The performance deficiency is more than minor because it adversely impacted the human performance attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the performance deficiency could have become more significant in that not completing the required simulator testing annually can lead to not detecting and correcting errors in the simulator so that it models the actual plant correctly. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, and the
 
corresponding Appendix I, Licensed Operator Requalification Significance Determination Process, the finding was determined to have very low safety significance (green) because there was no actual event caused by not modeling the actual plant correctly. This finding has no cross-cutting aspect assigned because the cause was not representative of current licensee performance.
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Assess and Manage Outage Risk Associated with Significant Switchyard Work The inspectors identified a non-cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, involving the licensees failure to assess and manage outage risk related to significant switchyard work. Specifically, the licensee allowed risk significant relay test work to result in loss of one of two offsite safety related 4 kV power feeds to the plant during Refueling Outage 18. With Callaway Plant in Mode 6, Refueling, the risk assessment for October 21, 2011, and the Outage Shutdown Management Plan prohibited significant switchyard work. However, at 1:21 p.m., emergency diesel generator A bus NB01 became deenergized due to improper switchyard testing. Callaway Action Request 201108888 was initiated to develop corrective actions.
Failure to properly assess and manage the risk of significant switchyard work during a high decay heat condition was a performance deficiency. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The offsite power system was affected by this event. Using Manual Chapter 0609, Appendix G, , Checklist 4 - PWR Refueling Operation: RCS level > 23 OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, this finding was determined to be of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory, did not degrade the ability to terminate a leak path or add reactor coolant system inventory when needed, and did not degrade Enclosure the ability to recover decay heat removal, if lost. This finding has a cross-cutting aspect in the area of human performance associated with the resources component because Procedure EDP-ZZ-01129, Callaway Plant Risk Assessment, Attachment 6, Step 6.c, was not sufficiently complete and accurate to define significant switchyard work.
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Improper Ground and Test Device Damages Residual Heat Removal Pump Switchgear The inspectors reviewed a Green self-revealing non-cited violation of 10 CFR Part 50 Appendix B, Criterion V, Procedures, involving the licensees failure to correctly install a ground test device for the train A safety-related 4160 volt switchgear, NB01. During maintenance on the train A safety related bus, workers improperly placed a ground test device with 2000 ampere stab adapters into the 1200 ampere breaker cubicle (for the residual heat removal pump). This damaged the switchgear connection point and caused the breaker to fail, rendering the pump inoperable.
The reactor was defueled so the residual heat removal system was not required by technical specifications at the time, but the bus was required to be removed from service for repairs. The licensee repaired the bus connection point, and the pump was retested satisfactorily. This finding was entered into the licensee's corrective action program as Callaway Action Request 201109122.
Failure to install the correctly configured ground and test device into the NB0101 cubicle of the NB01 switchgear was a performance deficiency. This is more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, improper maintenance caused the residual heat removal pump to become unavailable. Because no fuel was in the vessel at the time of the event, the inspectors referred the issue to a Region IV senior reactor analyst for the significance determination. The analyst used NRC Inspection Manual 0609, Appendix G, Shutdown Operations
 
Significance Determination Process, to evaluate the significance of the finding. Since all of the fuel had been removed from the vessel there was no potential for core damage (the delta core damage frequency was zero).
Therefore, the finding is of very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance associated with the resources component in that the licensee failed to ensure training of maintenance personnel was adequate to assure nuclear safety.
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Isolate Control Room Air Comditining Unite SGK04A for Maintenance The inspectors reviewed a Green self-revealing non-cited violation of Technical Specification 5.4.1.a, Procedures, involving the failure to isolate an electrical power supply during maintenance on control room air conditioning system train A. Specifically, while removing an electrical cabinet for maintenance, workers discovered an energized lead that was supposed to have been isolated for the work. Workers failed to stop work and make appropriate notifications. As a result, when the lead was reterminated, it grounded the bus and caused inverter NN11 to shift to an alternate power supply. This caused operators to make an unplanned entry into a 24-hour shutdown technical specification action statement. The licensee restored normal power to inverter NN11 within 4 hours. This issue was entered into the corrective action program as Callaway Action Request 201107612.
Failure to stop work when a lockout tagout isolation was discovered to be inadequate was a performance deficiency.
This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, inverter NN11 was rendered less reliable by the improper maintenance. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to stop in the face of uncertainty or unexpected circumstances.
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Implement Plant Maintenance Procdures The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a, Procedures, involving the failure to ensure compliance with relay test maintenance procedures associated with electrical switchyard work that affected the performance of safety related equipment. On October 21, 2011, Callaway Plant was in Mode 6 with switchyard activities in progress to test transfer trip and lockout relay devices. At 1:21 p.m. the control room operators received several annunciators indicating that diesel generator bus A and its safety related loads had become deenergized. An improperly operated lockout relay had cascaded a test signal onto other components in the plant electrical system. This issue was entered into the corrective action program as Callaway Action Request 201108691.
Failure to establish the safe working conditions per the transfer trip procedure and failure to operate the lockout relay in the manner specified by the lockout relay procedure were performance deficiencies. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, one of the two offsite power feeds to the plant was lost. Using Manual Chapter 0609 Appendix G Attachment 1, Checklist 4 - PWR Refueling Operation:
RCS level > 23 OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, this finding was determined to be of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory, did not degrade the ability to terminate a leak path or add reactor coolant system inventory when needed, and did not degrade the ability to recover decay heat removal. This finding has a cross-cutting
 
aspect in the area of human performance associated with the work controls component because the electrical relay test technicians, onsite engineering, and work control staff failed to adequately maintain interfaces to communicate and safely coordinate significant switchyard activities to ensure proper human performance.
Inspection Report# : 2011005 (pdf)
Significance:        Sep 23, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correctly Implement Plant Maintenance Procedures The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to correctly follow maintenance procedures which resulted in a failure of motor-operated valve EFHV0065 associated with the ultimate heat sink train A cooling tower. To perform its safety function the valve must be capable of being closed. On September 15, 2010, the mechanical maintenance department removed and rebuilt the actuator for the motor-operated valve. The valve actuator stop nuts were not set correctly and remained set outside the range of the electrical limits due to electrical maintenance workers failing to complete the procedure and work instructions initiated by the mechanical department. On June 22, 2011, an attempt to manually align essential service water return over the train A safety-related cooling tower failed when the motor-operated valve was manually positioned past the zero percent open position due to the improperly set stop nuts. This disengaged the valve operator worm from its worm gear, opened the valve, and rendered the valve being incapable of being closed. The immediate corrective action to replace the valve actuator was completed on June 23, 2011. The licensee initiated Callaway Action Request 201105074 to evaluate cause and extent-of-condition and specify corrective actions.
This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work controls component because the mechanical and electrical maintenance technicians failed to adequately maintain interfaces to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2011004 (pdf)
Significance:        Sep 23, 2011 Identified By: Self-Revealing Item Type: FIN Finding Failure to Evaluate Breaker Relay Settings Results in Partial Loss of Station Blackout Response Capability The inspectors reviewed a self-revealing finding for the failure of AmerenUE engineering personnel to correctly establish the relay settings for the alternate emergency power supply diesel output breakers. On August 21, 2011, Callaway Plant experienced a loss of power to the alternate emergency power supply diesel bus PA05. This resulted in all four alternate emergency power supply diesels starting; however, the number three diesel output breaker immediately tripped open. The licensee determined that the breakers protective relaying was improperly set. Further investigation by AmerenUE discovered that all four of the diesel output breakers had incorrect settings. The incorrect settings occurred due to the limited range of the relay chosen for the application and the engineering recommendations that prioritized protecting the diesel over limiting the margin to unintended breaker trips. Callaway engineering reviews had not identified the low margin to unintended trips. The licensee initiated corrective actions associated with Callaway Action Request 201106701 to change the differential current relay settings.
This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the finding is of very low safety significance because it was a design deficiency that did not result in a loss of system safety function, did not represent an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk-significant per 10 CFR 50.65, for greater than 24 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of problem identification and resolution because the
 
licensee failed to implement a corrective action program with a low threshold for identifying issues commensurate with their safety significance.
Inspection Report# : 2011004 (pdf)
Significance:      Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain an Adequate Flooding Analysis for Room 3101 The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after the licensee failed to provide adequate design control measures for verifying the adequacy of the flooding analysis associated with the 2009 modification that replaced essential service water carbon steel piping with high density polyethylene piping. The licensee did not update the flooding analysis of record to consider potential failures in the new piping. The licensee generated Callaway Action Request 201102957 to develop a means to evaluate the relative stresses associated with the new pipe.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding required a Phase 2 significance determination. Using the presolved worksheet from the Risk Informed Inspection Notebook for the Callaway Station, Revision 2.01a, the finding was red, which warranted further review. Therefore, a senior reactor analyst performed a bounding Phase 3 significance determination. The bounding change to the core damage frequency was approximately 4.1E-7 (Green). This was impacted significantly by the very small amount of new piping in the room.
This finding was determined to have a cross-cutting aspect in the area of Problem Identification and Resolution associated with the corrective action component in that the licensee did not thoroughly evaluate the extent of condition when the residents challenged the flooding calculation in December 2010 such that the resolutions addressed causes and extent of conditions, as necessary Inspection Report# : 2011003 (pdf)
Significance:      Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Analyze Refueling Water Storage Tank Level Transmitters for High-Energy Line Break The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to adequately evaluate a potential high-energy line break in nonseismically qualified auxiliary steam piping in the refueling water storage tank valve house. The harsh environment from a high-energy line break had the potential to impact safety related level transmitters associated with the refueling water storage tank. Following identification of this issue by the inspectors, the licensee analyzed the nonnuclear auxiliary piping to ensure it could withstand safe shutdown earthquake loadings which allowed high-energy line breaks at intermediate locations to be excluded. This issue was entered into the licensees corrective action program as Callaway Action Request 201102588.
This finding is greater than minor because it is associated with the Mitigating Systems Cornerstone attribute of design control and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 -
Initial Screening and Characterization of Findings, this finding is determined to be of very low safety significance since subsequent evaluation concluded the issue was a design or qualification deficiency confirmed not to result in loss of operability or functionality. This finding did not have a cross-cutting aspect since the error associated with the high-energy line break analysis was not reflective of current licensee performance.
Inspection Report# : 2011003 (pdf)
Significance:      Mar 24, 2011 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Document Reasonable Expectation of Operability for Equipment Supported b y the Class 1E Air Conditioning Units The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to adequately evaluate past operability associated with the Class 1E electrical equipment air conditioning unit. The inspectors identified that Revision 1 and 2 to Callaway Action Request 200800615 incorrectly concluded that the equipment supported by the Class 1E electrical equipment air conditioning unit train B was operable with the units cooling water flow control valve in manual. This issue was entered into the licensees corrective action program as Callaway Action Request 201102565.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of human performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened to a Phase 2 significance determination because it involved the loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a bounding Phase 3 significance determination and determined that the finding was of very low safety significance (Green). The very short exposure period coupled with the availability of train A equipment helped to mitigate the significance. The dominant core damage sequences included a loss of main feedwater initiating event; the loss of train B electrical power; and various failures of auxiliary feedwater. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee failed to use conservative assumptions including verifying the validity of the underlying assumptions when performing operability/reportability evaluations.
Inspection Report# : 2011002 (pdf)
Significance: SL-IV Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report Inoperability of Class IE Electrical Equipment for a Period Greater than Allowed by the Plant's Technical Specifications The inspectors identified a IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for failure to report inoperability of Class 1E electrical equipment for a period greater than allowed by the plants technical specifications. The licensee determined there were no prior instances where the Class 1E electrical equipment air conditioning units were inoperable greater than the technical specification allowed completion time of the supported equipment. The inspectors reviewed the licensees reportability evaluation and identified that the event described in Callaway Action Request 200800615 resulted in a period where the Class 1E electrical equipment air conditioning unit train B was inoperable for approximately 37 hours which exceeded the technical specification allowed completion time of the equipment supported by the Class 1E electrical equipment and constituted a condition which was prohibited by the plant's technical specifications and should have been reported in a licensee event report.
This issue was entered into the licensees corrective action program as Callaway Action Request 201011132.
This finding affects the Mitigating Systems Cornerstone and is greater than minor because in order to perform its regulatory function, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was strictly associated with a traditional enforcement violation.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Containment Spray Test Procedure Potentially Creates an Unanalyzed Condition The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to provide adequate procedural guidance for testing of containment spray pumps. The inspectors reviewed a licensee evaluation of the acceptability of their existing containment spray pump testing procedure and found that it failed to adequately address the underlying technical issues because it relied on operators
 
recognizing the diversion flow path and focused on the operability of the containment spray system and not the ability to maintain the long term cooling function of the emergency core cooling system. Additionally, the inspectors identified that the procedure would have provided a diversion flow path of post-accident sump fluids back to the refueling water storage tank exceeding those currently analyzed in the Callaway licensing bases. This issue was entered into the licensees corrective action program as Callaway Action Request 201011233 and the licensee implemented procedure changes to address the potential for post-loss of coolant accident containment sump fluids being injected back to the refueling water storage tank.
This finding was determined to be greater than minor because it impacted the Mitigating Systems Cornerstone attribute of procedure quality and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding screened to a Phase 2 significance determination because it involved a potential loss of safety function. A Region IV senior reactor analyst performed a bounding Phase 3 significance determination and determined that the finding was of very low safety significance (Green). The very short exposure period coupled with the availability of equipment needed for other initiating events (other than small and medium loss of coolant accidents) helped to mitigate the significance. The dominant core damage sequences included small and medium break loss of coolant accidents, and the failure of emergency core cooling pumps in the recirculation mode. This finding was determined not to have a crosscutting aspect since the performance deficiency is not reflective of current performance.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 24, 2011 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding Installation Inadequacy The inspectors identified a noncited violation of Technical Specification 5.4.1.a for failure to properly implement Procedure MDP-ZZ-S0001, "Scaffolding Installation and Evaluation," Revision 26, when scaffolding was erected near or in contact with equipment in safety-related structures. On February 8 and March 16, 2011, the inspectors identified two locations where scaffold poles and a scaffold pin were less than the procedure required 1 inch from the auxiliary building vent line, the Train B emergency diesel lube oil drain line, and also essential service water system piping in the Train B diesel room. This issue was entered into the licensees corrective action program as Callaway Action Request 201102091.
The deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding was associated with the Mitigating Systems Cornerstone. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the component of corrective action program because the licensee did not have a low threshold for identifying scaffold issues.
Inspection Report# : 2011002 (pdf)
Significance: SL-IV Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report The team identified a Severity Level IV, noncited violation of 10 CFR 50.71, Maintenance of records, making of reports, paragraph (e) which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the updated safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Specifically, the licensee incorporated numerous errors in the updated safety analysis report associated with the descriptions of the onsite electrical power systems. The licensee has entered this violation into their corrective action program as Condition Reports 201101335 and 201102064.
 
The inspectors determined that the failure to update the updated safety analysis report as required by 10 CFR 50.71(e),
Maintenance of records, making of reports was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function.
The inspectors used the NRC Enforcement Policy, dated September 30, 2010, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect as it was associated with a traditional enforcement violation.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Design the Emergency Diesel Generator Ground Fault Protection Circuitry The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, when designing the bypass circuitry for the emergency diesel generator ground fault trip function, the licensee failed to ensure that the associated electrical components were adequately designed for the continuous duty they would have to withstand under bypassed trip conditions. This could result in an ignition source and subsequent fire in the area under these conditions. This finding was entered into the licensees corrective action program as Condition Report 201102064.
The team determined that the failure to analyze the suitability of the emergency diesel generator components when protection features were bypassed was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design of these components could have prevented continued operation of the emergency diesel generator under ground fault conditions with the trip signal bypassed. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Specifically, the licensee revised the associated procedures to include these components in the combustible material exclusion zone. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Residual Heat Removal Flow Alarm Setpoint The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of March 3, 2011, the Mode 6 residual heat removal system low flow alarm setpoint did not adequately account for flow measurement uncertainties, and consequently was non-conservative. The licensee has entered the violation into their corrective action program as Condition Report 201101750.
The team determined that the failure to adequately analyze the uncertainty in measurement of residual heat removal system flow, and the impact of this failure, was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design basis analysis, and plant instrumentation, did not ensure that, while operating in Mode 6, the control room operators would be alerted whenever the residual heat removal system flow through the reactor coolant system was below the required value of 1000 gallons per minute. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the
 
issue was determined to have very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011006 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure in the Establishment of a Turbine Driven Auxiliary Feedwater Pump Postmaintenance Test Procedure The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving a failure to follow procedures in the development of Procedure OTS-FC-0006, "TDAFW Pump Post-Maintenance Test Run on Aux Steam." Specifically, the licensee failed to incorporate turbine-driven auxiliary feedwater pump vendor manual precautions, limitations, and technical information in Procedure OTS-FC-0006, which resulted in the axial unloading, rolling element ball skidding, and subsequent degradation to the turbine-driven auxiliary feedwater pump inner outboard thrust bearing. Following discovery during planned maintenance and as immediate corrective actions, the licensee declared the turbine-driven auxiliary feedwater pump inoperable, entered the applicable Technical Specification Limiting Condition for Operation, replaced the oil and bearings, restored the pump to operability, and initiated Callaway Action Request 201101042 to perform a root cause analysis.
This finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and
-~----------eh-a-ra-cterizationofFfhalngs," and was determined to be of very low safety significance (Green) because there was not a design or qualification deficiency that resulted in a loss of operability or functionality, it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time, it did not represent an actual loss of risk significant equipment, and it did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to ensure procedural adherence in the establishment of the turbine-driven auxiliary feedwater pump postmaintenance test procedure.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure for Turbine-Driven Auxiliary Feedwater Pump Postmaintenance Testing The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving five examples of failure to follow 2 Enclosure Procedure OTS-FC-0006, "TDAFW Pump Post-Maintenance Test Run on Aux Steam." Specifically, operators failed to follow an existing total flow precaution in Procedure OTS-FC-0006 which resulted in the axial unloading, rolling element ball skidding, and subsequent degradation to the turbine-driven auxiliary feedwater pump inner outboard thrust bearing. Following initial condition discovery during planned maintenance and as immediate corrective actions, the
 
licensee declared the turbine-driven auxiliary feedwater pump inoperable, entered the applicable Technical Specification Limiting Condition for Operation, replaced the oil and bearings, restored the pump to operability, and initiated Callaway Action Request 201101042 to perform a root cause analysis.
These findings were more than minor because they affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding was evaluated Using Manual Chapter 0609.04, "Phase 1 -Initial Screening and Characterization of Findings." These findings were determined to be of very low safety significance (Green) because there was not a design or qualification deficiency that resulted in a loss of operability or functionality, they did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time, they did not represent an actual loss of risk significant equipment, and they did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to ensure procedural adherence in the implementation of the turbine-driven auxiliary feedwater pump postmaintenance test procedure.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Past Operability Determination The inspectors identified a non cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," regarding the licensee's failure to follow the requirements of Procedure APA-ZZ-00500, Appendix 3, "Past Operability and Reportability Evaluations." Specifically, the inspectors identified that the past operability evaluation for the turbine-driven auxiliary feedwater pump used a nonconservative calculation of mission time that did not take into account all design and licensing basis functions when determining the mission time. The licensee entered this issue into their corrective action program as Callaway Action Request 201102431 and updated its mission time anal}"sis to account for the turbine-driven auxiliary feedwater pump's specified safety function to bring the plant to a safe shutdown condition.
This finding is greater than minor because if left uncorrected, it would have the potential to lead to a more Significant safety concern because systems that may be inoperable may not be recognized and that it impacted the Mitigating Systems Cornerstone attribute of human performance in that the failure to accurately understand the auxiliary feedwater system mission time affected the mitigating systems objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, "Phase 1 -Initial Screening and Characterization of Findings," the finding is determined to have very low safety significance because it did not result in the loss of safety function of any technical specification required equipment. The cause of this finding is related to the problem identification and resolution cross-cutting component of corrective action program because licensee personnel failed to thoroughly evaluate conditions adverse to quality and perform adequate operability determinations.
Inspection Report# : 2011007 (pdf)
Significance:        Feb 08, 2011 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Calculate and Implement Conservative Safety Related Equipment Oil Leakage Operability Criteria The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for a failure to adequately determine safety related equipment oil leakage acceptance criteria that was used in operator logs. Specifically, the 2008 licensee fluid leak management program calculations to determine the mission time assessments related to oil leak rates of safety related pumps and motors were nonconservative when added to Procedure ODP-ZZ-0016E, Appendix 1, "Equipment Operator General Inspection Guide." The licensee evaluated this issue in Callaway Action Request 201102431 and calculated new conservative oil leak rates for the affected equipment.
This finding is more than minor because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to adequately evaluate and determine an appropriate lube oil leak rate to maintain safety related equipment operability affects the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings,"
and determined to be of very low safety significance since the as-found condition of the safety related equipment reviewed back to August 15, 2007, found no oil leakage rates that would have caused a loss of system safety function. This finding was not reflective of current licensee performance and therefore, has no cross-cutting aspect.
Inspection Report# : 2011007 (pdf)
Significance:      Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance Schedule to Protect Safety-Related Equipment from Undetected Degraded Conditions The inspectors identified a noncited violation of Technical Specification 5.4.1 for a failure to adequately establish and implement procedures required by Regulatory Guide 1.33, Appendix A, Section 9, "Procedures for Performing Maintenance." Specifically, the preventative maintenance schedule to perform periodic lube oil analysis established an 18 month frequency without adequate justification resulting in a failure to promptly detect bearing degradation in the turbine-driven auxiliary feedwater pump. The licensee evaluated this issue in Callaway Action Request 201101042 and has corrective actions to review the lube oil analysis frequency and reduce it to at least a 9 month frequency-,-. _______________ _
This finding is more than minor because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to adequately evaluate and determine an appropriate lube oil monitoring schedule resulted in the failure to promptly detect a degraded bearing in the turbine-driven auxiliary feedwater pump affecting the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. The finding was evaluated using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings,"
and determined to be of very low safety significance since the as-found condition of the degraded bearing would not have caused a loss of system safety function.
The finding has a cross-cutting aspect in the area of human performance associated with the decision making component, in that, the licensee failed to use conservative assumptions in the decision to extend the turbine-driven auxiliary feedwater pump lube oil monitoring interval to 18 months.
Inspection Report# : 2011007 (pdf)
 
Significance:        Feb 08, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Cause Results in Missed Safety-Related Pump Overhaul The inspectors identified a noncited violation of 10 CFR Part 50, Appendix 8, Criterion XVI, "Corrective Action," associated with the licensee's failure to promptly identify and correct a condition adverse to quality.
Specifically, the licensee reduced the scope of preventative maintenance for the turbine-driven auxiliary feedwater pump overhaul during Refueling Outage 16 without proper justification, resulting in the failure to perform required pump maintenance. This issue was entered into the licensee's corrective action program as Callaway Action Request 201102407 and the pump has been scheduled to be overhauled during the next refueling outage.
This finding is more than minor because, if left uncorrected, corrective action deficiencies would have the potential to lead to a more significant safety concern.
The failure to perform required maintenance could allow equipment degradation affecting the equipment performance attribute of the Mitigating Systems Cornerstone and could have impacted the availability of mitigating equipment if left uncorrected. Using Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," this finding was determined to be of very low safety significance because the degraded condition did not result in a loss of operability or functionality. The inspectors determined that the finding has a cross-cutting aspect in the area of human performance associated with the work control component because the licensee does not appropriately coordinate work activities by incorporating actions to address the impact of changes to work scope on the plant such that nuclear safety is supported.
Inspection Report# : 2011007 (pdf)
Barrier Integrity Significance:        Jun 23, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Establish Test Program for Isolation Valves in Post-LOCA Recirculation Flowpath The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for failure to adequately demonstrate that the seat leakage of centrifugal charging pump and safety injection pump suction isolation valves remained within acceptable limits. These valves have a combined allowable leakage rate of three gallons per minute to ensure that offsite thyroid and whole body doses remain within regulatory limits. Since the flowpaths have isolation valves for which seat leakage is limited to a specific maximum amount, the inspectors identified that they should be considered Category A valves as specified in ASME OM Code which requires the valves be tested at least once every two years. At the end of the inspection period, the licensee was planning a recurring surveillance test to verify seat leakage for these valves is within acceptable limits. This issue was entered into the licensees corrective action program as Callaway Action Request 201104577.
This finding was greater than minor because it was associated with the Barrier Integrity Cornerstone attribute of configuration control and affects the associated cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the issue was determined to represent an actual open pathway in the physical integrity of reactor containment. Using Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, this finding was determined to be a Type B finding since it was related to a degraded condition that has potentially important implications for the integrity of containment, without affecting the likelihood of core damage. This finding was found to be of very low safety significance since the nontested flowpath would be comparable to small lines (less
 
than 1 2 inches in diameter) and would not generally contribute to large early release frequency. This finding did not have a cross-cutting aspect since the error associated with the inservice testing program was not reflective of current licensee performance.
Inspection Report# : 2011003 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:        May 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Recipients License Conditions Prior to shipping Special Nuclear Material The inspectors identified a noncited violation of 10 CFR 70.42 (c) for failure to verify that a recipient of special nuclear material was authorized to receive the quantity of material shipped. This finding was determined to be of very low safety significance. Specifically, On June 15, 2010, the licensee shipped laundry contaminated with radioactive material to a state licensed processing facility in Alabama. The licensee verified that the processing facility was licensed to handle the material being shipped, but failed to verify that the recipients license authorized the quantity of material shipped. The licensee notified the Alabama licensee and proposed a revision to the shipping procedures. This violation was entered into the licensees corrective action program as Callaway Action Request 201104385.
This finding was greater than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process (transportation program), and affected the cornerstone objective, in that, license conditions were violated and these conditions are in place, in part, to control exposure to radiation. Using the public radiation safety significance determination process, the inspectors determined the finding had very low safety significance because (1) radiation limits were not exceeded, (2) there was no breach of a package during transit, (3) it did not involve a certificate of compliance issue, (4) it was not a low level burial ground nonconformance, and (5) it did not involve a failure to make notifications or provide emergency information. This finding had a crosscutting aspect in the area of human performance, resources component, because licensee procedures were inadequate to ensure proper shipping of radioactive material and that license conditions were not violated.
Inspection Report# : 2011003 (pdf)
Significance: SL-IV May 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Periodically Update the Final Safety Analysis Report The inspectors identified a noncited violation of 10 CFR 50.71 Maintenance of Records, because the licensee failed to update their Final Safety Analysis Report with submittals that include the effects of a change made to the facility.
Specifically, the licensee built the old steam generator storage facility on the owner controlled area for long-term radwaste storage of four decommissioned steam generators and failed to update the Final Safety Analysis Report to include these changes to the facility. This issue was entered in the licensees corrective action program as Callaway Action Request 201104434.
This issue was dispositioned using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding is more than minor because it has a material impact on licensed activities in that the four decommissioned steam generators, with a significant radioactive source term, have been relocated from the plant radiological controlled area to the owner controlled area. In addition, the radwaste
 
management program has been affected because the licensee determined that this low-level radwaste facility will store these large components until an appropriate facility for disposal can be determined. The finding is characterized as a Severity Level IV noncited violation in accordance with NRC Enforcement Policy, Section 6.1, and was treated as a noncited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy.
Inspection Report# : 2011003 (pdf)
Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : March 02, 2012
 
Callaway 1Q/2012 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Separation of Stainless Steel and Carbon Steel Hand Files and Wire Brushes The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, for the failure to have procedures that ensured that hand files and wire brushes designated for stainless steel weld preparation were stored separately from hand files and wire brushes used on carbon steel. The licensee took corrective actions to remove the stainless steel designations from stainless steel tools that were mixed with tools used on carbon steel, established segregated locations in tool rooms for the separation of abrasive tools, and trained tool room attendants to properly store and mark abrasive tools designated for use on stainless steel. This issue was entered into the licensees corrective action program as Callaway Action Request 201108921.
Inspectors determined that the failure to assure that hand files and wire brushes designated for exclu}}

Latest revision as of 13:52, 29 November 2024

2017 Q1-Q4 ROP Inspection Findings
ML20311A628
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Issue date: 11/06/2017
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