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{{#Wiki_filter:1Q/2000 Inspection Findings - River Bend 1                                                                                                Page 1 of 12 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
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The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:        Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:        Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:        Jul 11, 2001
 
1Q/2000 Inspection Findings - River Bend 1                                                                                            Page 2 of 12 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system.
Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas.
The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire
 
1Q/2000 Inspection Findings - River Bend 1                                                                                              Page 3 of 12 Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event.
The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and
 
1Q/2000 Inspection Findings - River Bend 1                                                                                              Page 4 of 12 assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure.
Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval.
 
1Q/2000 Inspection Findings - River Bend 1                                                                                            Page 5 of 12 During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions.
Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:        Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective
 
1Q/2000 Inspection Findings - River Bend 1                                                                                              Page 6 of 12 actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Inspection Report# : 2000016(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages.
The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee
 
1Q/2000 Inspection Findings - River Bend 1                                                                                              Page 7 of 12 Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
 
1Q/2000 Inspection Findings - River Bend 1                                                                                            Page 8 of 12 Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Barrier Integrity Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure
 
1Q/2000 Inspection Findings - River Bend 1                                                                                            Page 9 of 12 RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys,"
requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed.
Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Significance:        Jun 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2000 Inspection Findings - River Bend 1                                                                                              Page 10 of 12 Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions.
Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters.
Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection
 
1Q/2000 Inspection Findings - River Bend 1                                                                                            Page 11 of 12 Significance:          Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:          May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 18, 2000 Identified By: NRC
 
1Q/2000 Inspection Findings - River Bend 1                                                                                              Page 12 of 12 Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Last modified : April 01, 2002
 
2Q/2000 Inspection Findings - River Bend 1                                                                                              Page 1 of 12 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages.
The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification
 
2Q/2000 Inspection Findings - River Bend 1                                                                                              Page 2 of 12 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time
 
2Q/2000 Inspection Findings - River Bend 1                                                                                                Page 3 of 12 testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Significance:        Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:        Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:        Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
 
2Q/2000 Inspection Findings - River Bend 1                                                                                            Page 4 of 12 Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system.
Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas.
The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner
 
2Q/2000 Inspection Findings - River Bend 1                                                                                              Page 5 of 12 sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event.
The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
 
2Q/2000 Inspection Findings - River Bend 1                                                                                              Page 6 of 12 Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure.
Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval.
During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000
 
2Q/2000 Inspection Findings - River Bend 1                                                                                            Page 7 of 12 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions.
Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:        Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently,
 
2Q/2000 Inspection Findings - River Bend 1                                                                                              Page 8 of 12 two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Inspection Report# : 2000016(pdf)
Barrier Integrity Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders
 
2Q/2000 Inspection Findings - River Bend 1                                                                                            Page 9 of 12 Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys,"
requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed.
Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Significance:        Jun 23, 2001 Identified By: NRC
 
2Q/2000 Inspection Findings - River Bend 1                                                                                              Page 10 of 12 Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions.
Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters.
Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection
 
2Q/2000 Inspection Findings - River Bend 1                                                                                            Page 11 of 12 Significance:          Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:          May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 18, 2000 Identified By: NRC
 
2Q/2000 Inspection Findings - River Bend 1                                                                                              Page 12 of 12 Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Last modified : April 01, 2002
 
3Q/2000 Inspection Findings - River Bend 1                                                                                              Page 1 of 12 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The
 
3Q/2000 Inspection Findings - River Bend 1                                                                                              Page 2 of 12 safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000016(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages.
The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:        May 06, 2000 Identified By: NRC
 
3Q/2000 Inspection Findings - River Bend 1                                                                                            Page 3 of 12 Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee
 
3Q/2000 Inspection Findings - River Bend 1                                                                                                Page 4 of 12 subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Significance:        Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:        Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:        Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
 
3Q/2000 Inspection Findings - River Bend 1                                                                                            Page 5 of 12 Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system.
Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas.
The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event.
The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse
 
3Q/2000 Inspection Findings - River Bend 1                                                                                              Page 6 of 12 to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure.
Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
 
3Q/2000 Inspection Findings - River Bend 1                                                                                            Page 7 of 12 Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval.
During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
 
3Q/2000 Inspection Findings - River Bend 1                                                                                            Page 8 of 12 Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions.
Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Barrier Integrity Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
 
3Q/2000 Inspection Findings - River Bend 1                                                                                            Page 9 of 12 Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys,"
requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed.
Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
 
3Q/2000 Inspection Findings - River Bend 1                                                                                              Page 10 of 12 Significance:        Jun 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions.
Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters.
Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
 
3Q/2000 Inspection Findings - River Bend 1                                                                                            Page 11 of 12 Physical Protection Significance:          Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:          May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
 
3Q/2000 Inspection Findings - River Bend 1                                                                                              Page 12 of 12 Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Last modified : March 29, 2002
 
4Q/2000 Inspection Findings - River Bend 1                                                                                              Page 1 of 12 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use
 
4Q/2000 Inspection Findings - River Bend 1                                                                                              Page 2 of 12 members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure.
Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval.
During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with
 
4Q/2000 Inspection Findings - River Bend 1                                                                                            Page 3 of 12 Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions.
Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:        Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
 
4Q/2000 Inspection Findings - River Bend 1                                                                                              Page 4 of 12 Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Inspection Report# : 2000016(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages
 
4Q/2000 Inspection Findings - River Bend 1                                                                                            Page 5 of 12 The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages.
The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact
 
4Q/2000 Inspection Findings - River Bend 1                                                                                                Page 6 of 12 implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:        Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:        Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:        Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
 
4Q/2000 Inspection Findings - River Bend 1                                                                                            Page 7 of 12 Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system.
Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas.
The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner
 
4Q/2000 Inspection Findings - River Bend 1                                                                                            Page 8 of 12 sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event.
The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Barrier Integrity Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected
 
4Q/2000 Inspection Findings - River Bend 1                                                                                            Page 9 of 12 and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys,"
requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed.
Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Significance:        Jun 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions
 
4Q/2000 Inspection Findings - River Bend 1                                                                                              Page 10 of 12 The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions.
Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters.
Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Physical Protection
 
4Q/2000 Inspection Findings - River Bend 1                                                                                            Page 11 of 12 Significance:          Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:          May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or
 
4Q/2000 Inspection Findings - River Bend 1                                                                                              Page 12 of 12 credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Last modified : March 28, 2002
 
1Q/2001 Inspection Findings - River Bend 1                                                                                              Page 1 of 12 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system.
Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The
 
1Q/2001 Inspection Findings - River Bend 1                                                                                            Page 2 of 12 inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas.
The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event.
The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC
 
1Q/2001 Inspection Findings - River Bend 1                                                                                              Page 3 of 12 Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure.
Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding
 
1Q/2001 Inspection Findings - River Bend 1                                                                                            Page 4 of 12 Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval.
During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions.
Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not
 
1Q/2001 Inspection Findings - River Bend 1                                                                                              Page 5 of 12 performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:        Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Inspection Report# : 2000016(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
 
1Q/2001 Inspection Findings - River Bend 1                                                                                              Page 6 of 12 Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages.
The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
 
1Q/2001 Inspection Findings - River Bend 1                                                                                                Page 7 of 12 Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:          Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:          Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
 
1Q/2001 Inspection Findings - River Bend 1                                                                                              Page 8 of 12 The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:        Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Barrier Integrity Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders
 
1Q/2001 Inspection Findings - River Bend 1                                                                                            Page 9 of 12 Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys,"
requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed.
Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Significance:        Jun 23, 2001 Identified By: NRC
 
1Q/2001 Inspection Findings - River Bend 1                                                                                              Page 10 of 12 Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions.
Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters.
Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection
 
1Q/2001 Inspection Findings - River Bend 1                                                                                            Page 11 of 12 Significance:          Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:          May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or
 
1Q/2001 Inspection Findings - River Bend 1                                                                                              Page 12 of 12 credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Last modified : March 28, 2002
 
2Q/2001 Inspection Findings - River Bend 1                                                                                              Page 1 of 12 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system.
Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The
 
2Q/2001 Inspection Findings - River Bend 1                                                                                            Page 2 of 12 inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas.
The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event.
The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC
 
2Q/2001 Inspection Findings - River Bend 1                                                                                              Page 3 of 12 Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure.
Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding
 
2Q/2001 Inspection Findings - River Bend 1                                                                                            Page 4 of 12 Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval.
During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations
 
2Q/2001 Inspection Findings - River Bend 1                                                                                                Page 5 of 12 determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions.
Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:        Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Inspection Report# : 2000016(pdf)
Significance:        Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not
 
2Q/2001 Inspection Findings - River Bend 1                                                                                              Page 6 of 12 represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:        Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:        Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
 
2Q/2001 Inspection Findings - River Bend 1                                                                                          Page 7 of 12 Significance:          Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages.
The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
 
2Q/2001 Inspection Findings - River Bend 1                                                                                            Page 8 of 12 Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Barrier Integrity Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure
 
2Q/2001 Inspection Findings - River Bend 1                                                                                                Page 9 of 12 RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Jun 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions.
Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys,"
requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
 
2Q/2001 Inspection Findings - River Bend 1                                                                                            Page 10 of 12 The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed.
Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters.
Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection
 
2Q/2001 Inspection Findings - River Bend 1                                                                                            Page 11 of 12 Significance:          Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:          May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or
 
2Q/2001 Inspection Findings - River Bend 1                                                                                              Page 12 of 12 credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Last modified : March 27, 2002
 
3Q/2001 Inspection Findings - River Bend 1                                                                                              Page 1 of 12 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:        Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:        Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive
 
3Q/2001 Inspection Findings - River Bend 1                                                                                            Page 2 of 12 maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system.
Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas.
The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event.
The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
 
3Q/2001 Inspection Findings - River Bend 1                                                                                              Page 3 of 12 Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule
 
3Q/2001 Inspection Findings - River Bend 1                                                                                              Page 4 of 12 The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure.
Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval.
During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical
 
3Q/2001 Inspection Findings - River Bend 1                                                                                                Page 5 of 12 deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions.
Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:        Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of
 
3Q/2001 Inspection Findings - River Bend 1                                                                                              Page 6 of 12 portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000016(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000
 
3Q/2001 Inspection Findings - River Bend 1                                                                                              Page 7 of 12 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages.
The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:          Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
 
3Q/2001 Inspection Findings - River Bend 1                                                                                            Page 8 of 12 Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Barrier Integrity Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-
 
3Q/2001 Inspection Findings - River Bend 1                                                                                                Page 9 of 12 2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed.
Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Significance:        Jun 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions.
Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys,"
 
3Q/2001 Inspection Findings - River Bend 1                                                                                            Page 10 of 12 requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters.
Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in
 
3Q/2001 Inspection Findings - River Bend 1                                                                                            Page 11 of 12 Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection Significance:          Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:          May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend
 
3Q/2001 Inspection Findings - River Bend 1                                                                                              Page 12 of 12 The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Last modified : March 26, 2002
 
4Q/2001 Inspection Findings - River Bend 1                                                                                                Page 1 of 11 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:        Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:        Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:        Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that
 
4Q/2001 Inspection Findings - River Bend 1                                                                                            Page 2 of 11 significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system.
Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas.
The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event.
The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
 
4Q/2001 Inspection Findings - River Bend 1                                                                                              Page 3 of 11 This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure.
Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC
 
4Q/2001 Inspection Findings - River Bend 1                                                                                            Page 4 of 11 Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval.
During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
 
4Q/2001 Inspection Findings - River Bend 1                                                                                              Page 5 of 11 Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions.
Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:        Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Inspection Report# : 2000016(pdf)
Significance:        Jun 24, 2000
 
4Q/2001 Inspection Findings - River Bend 1                                                                                              Page 6 of 11 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages.
The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope
 
4Q/2001 Inspection Findings - River Bend 1                                                                                            Page 7 of 11 The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:          May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Barrier Integrity Significance:          May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:          Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing
 
4Q/2001 Inspection Findings - River Bend 1                                                                                            Page 8 of 11 apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys,"
requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed.
Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
 
4Q/2001 Inspection Findings - River Bend 1                                                                                                Page 9 of 11 Significance:        Jun 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions.
Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters.
Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection
 
4Q/2001 Inspection Findings - River Bend 1                                                                                            Page 10 of 11 Significance:          Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:          May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent
 
4Q/2001 Inspection Findings - River Bend 1                                                                                              Page 11 of 11 licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Last modified : March 01, 2002
 
1Q/2002 Inspection Findings - River Bend 1                                                                                  Page 1 of 13 River Bend 1 Initiating Events Significance:        Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:        Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid.
Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:        Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of  to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
 
1Q/2002 Inspection Findings - River Bend 1                                                                                Page 2 of 13 Significance:        Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system. Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown
 
1Q/2002 Inspection Findings - River Bend 1                                                                              Page 3 of 13 equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas. The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event. The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1). This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing.
Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy. This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment. Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000
 
1Q/2002 Inspection Findings - River Bend 1                                                                                Page 4 of 13 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills. The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:          Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:          Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure. Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)
(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:          Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04. The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:          Oct 04, 2000 Identified By: NRC
 
1Q/2002 Inspection Findings - River Bend 1                                                                                Page 5 of 13 Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval. During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that: operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B. Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:        Aug 05, 2000
 
1Q/2002 Inspection Findings - River Bend 1                                                                                Page 6 of 13 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked. Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed.
Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009. This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Inspection Report# : 2000016(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages. The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action
 
1Q/2002 Inspection Findings - River Bend 1                                                                              Page 7 of 13 program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system.
The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
1Q/2002 Inspection Findings - River Bend 1                                                                              Page 8 of 13 Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Barrier Integrity Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance: TBD Mar 20, 2002 Identified By: NRC Item Type: AV Apparent Violation Failure to develop a range of protective actions, disseminate emergency response information, and maintain the emergency plan for members of the public located within the owner controlled area The inspector identified one preliminary finding involving the failure to develop a range of protective actions, disseminate emergency response information, and maintain the emergency plan in accordance with the requirements of 10 CFR 50.54(q), planning standards §50.47(b)(10) and (7), and 10 CFR Part 50, Appendix E, section IV(G) pertaining to members of the public located in the owner controlled area. Three apparent violations are associated with the finding. The issues involved: (1) a failure to establish effective means or provisions for warning, advising, evacuating, and monitoring members of the public during an owner controlled area evacuation, (2) a failure to disseminate emergency response information to the public using facilities in the River Bend Station owner controlled area, and (3) a failure to update the emergency plan and procedures after the public was permitted access to facilities in the owner controlled area. The licensee has entered these issues into its corrective action program in CR-RBS-2001-1713 and CR-RBS-2002-0183. This issue was preliminarily determined to have substantial safety significance (Yellow) because it
 
1Q/2002 Inspection Findings - River Bend 1                                                                              Page 9 of 13 represented a failure to meet a risk-significant emergency preparedness planning standard.
Inspection Report# : 2002005(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A. Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys," requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. This violation is being
 
1Q/2002 Inspection Findings - River Bend 1                                                                                Page 10 of 13 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inaudible alarm for personal electronic dosimeter used in a high radiation area Technical Specification 5.7.1.b states, in part, that any individual or group of individuals permitted to enter a high radiation area shall be provided with a radiation monitoring device that continuously integrates the radiation dose rate and alarms when a preset integrated dose is received. On October 5, 2001, the licensee identified that an individual working in a high radiation area was unable to hear his electronic dosimeter alarming on the dose accumulated alarm. Because the individual was unable to respond to the aural alarm, the device was inadequate to fulfill its Technical Specification required function. This violation is being treated as a noncited violation and is in the licensee's corrective action program as CR-RBS-2001-1325. 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, no substantial potential for overexposure, and no impact on the ability to assess dose.
Inspection Report# : 2001007(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed. Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Significance:        Jun 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions. Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials
 
1Q/2002 Inspection Findings - River Bend 1                                                                                Page 11 of 13 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards.
On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls. Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -
0982).
Inspection Report# : 2000016(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters. Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy.
This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection Significance:        Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:        May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available
 
1Q/2002 Inspection Findings - River Bend 1                                                                                Page 12 of 13 in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation." The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards.
Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000).
Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC
 
1Q/2002 Inspection Findings - River Bend 1                                                                                  Page 13 of 13 requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Last modified : July 22, 2002
 
2Q/2002 Inspection Findings - River Bend 1                                                                      Page 1 of 19 River Bend 1 Initiating Events Significance:      Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:      May 29, 2002 Identified By: NRC Item Type: FIN Finding Increased Division I Emergency Diesel Generator jacket cooling water leak rate caused diesel generator to be operable but degraded beyond the licensee's existing evaluation Following maintenance performed on May 9, 2002, to determine the source of a leak from the Division 1 emergency diesel generator jacket cooling water system, the leak rate more than doubled. The licensee's attempt to correct the problem on May 30, 2002, resulted in another increase in the leak rate to the point that makeup to the jacket cooling water system would be required within approximately 2 hours of Division I emergency diesel generator operation during a loss of offsite power. Although, the cause for the increased jacket water leak was repaired on June 4, 2002, the diesel generator remained degraded, but operable. The licensee planned to repair the original leak during the next extended diesel generator maintenance outage. The inspectors determined that the increased leak rate was beyond the licensee's evaluation that concluded that the Division 1 emergency diesel generator was degraded but operable. If left uncorrected, the jacket cooling water leak could have caused the emergency diesel generator to become inoperable and unavailable. The normal source of makeup water would not have been available during a loss of offsite power and the licensee did not develop a written procedure for use of an alternate makeup source until May 30, 2002. Using the significance determination process, the risk significance of the finding was determined to be very low because the emergency diesel generator remained operable, although degraded. This maintenance induced problem was documented in the licensee's corrective action program as Condition Report CR-RBS-2002-0672.
Inspection Report# : 2002002(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                      Page 2 of 19 Significance:      May 12, 2002 Identified By: Self Disclosing Item Type: FIN Finding Operator action caused a high reactor water level trip of the running reactor feed pump following a planned scram from 26 percent power Following a planned reactor scram during a plant shutdown, operators failed to take manual control of the feedwater level control system in time to stop an unexpected rise in reactor water level until after the running reactor feed pump tripped on high reactor water level. The licensee determined that the reduction of the reactor pressure control setpoint and subsequent opening of the main turbine bypass valves caused a "swell" in reactor water level which contributed to the higher than expected reactor water level transient. The inspectors determined that the operators did not manually close and isolate one of the two automatic feedwater regulating valves in time to eliminate leakage past the feedwater regulating valve, and failed to reject water from the reactor through the reactor water cleanup system in time to stop the rise in reactor water level to the high level trip of the reactor feed pump. The failure of the operators to manually control reactor water level resulted in the unavailability of a risk-significant reactor feed pump. The inspectors, using the significance determination process, determined that the safety significance of the high reactor water level trip of the running reactor feed pump following a planned reactor scram was very low because the reactor feed pump was restarted from the main control room as soon as reactor water level was lowered and the high reactor water level trip signal was cleared, and other reactor water makeup sources remained available. This human performance error was documented in the licensee's corrective action program as Condition Report CR-RBS-2002-0688.
Inspection Report# : 2002002(pdf)
Significance:      May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding Station Blackout Diesel Generator inoperable due to discharged starting battery The station blackout diesel generator was found to be inoperable by the licensee because its starting battery had been allowed to completely discharge. The station blackout diesel generator had been moved from its normal storage location as a contingency for a planned maintenance outage on several Division I safety-related systems. The inspectors determined that the Division I maintenance outage contingency plan and the weekly work schedule did not plan for the return of the station blackout diesel generator to its normal storage location to re-energize its battery charger. The licensee determined that this is a repeat of a similar event of April 4, 1998, documented in Condition Report CR-RBS-1998-0384. The failure to maintain its starting battery charged caused the risk significant station blackout diesel generator to be inoperable and unavailable. The inspectors, using the significance determination process, determined that the safety significance of the unavailability of the station blackout diesel generator was very low because the length of time the diesel generator was unavailable was less than 24 hours and all other electrical systems were available during that time. This human performance error was documented is the licensee's corrective action program as Condition Report CR-RBS-2002-0664.
Inspection Report# : 2002002(pdf)
Significance:      Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                      Page 3 of 19 instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:      Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:      Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Significance:      Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 4 of 19 perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system. Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                        07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                  Page 5 of 19 Significance:      Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas. The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:      Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event. The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy.
This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:      Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment.
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2Q/2002 Inspection Findings - River Bend 1                                                                      Page 6 of 19 Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills.
The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure. Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 7 of 19 relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04.
The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval. During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that:
file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 8 of 19 operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:      Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B.
Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:      Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                        07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 9 of 19 Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
Significance:      Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009.
This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000011(pdf)
Inspection Report# : 2000016(pdf)
Significance:      Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 10 of 19 of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages. The failure to identify the appropriate postmaintenance testing requirements as required by planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 11 of 19 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                  Page 12 of 19 closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Barrier Integrity Significance:      May 06, 2000 Identified By: NRC Item Type: FIN Finding Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:      Mar 20, 2002 Identified By: NRC Item Type: VIO Violation Failure to develop a range of protective actions, disseminate emergency response information, and maintain the emergency plan for members of the public located within the owner controlled area The inspector identified one preliminary finding involving the failure to develop a range of protective actions, disseminate emergency response information, and maintain the emergency plan in accordance with the requirements of 10 CFR 50.54(q), planning standards §50.47(b)(10) and (7), and 10 CFR Part 50, Appendix E, section IV(G) pertaining to members of the public located in the owner controlled area. Three apparent violations are associated with the finding. The issues involved: (1) a failure to establish effective means or provisions for warning, advising, evacuating, and monitoring members of the public during an owner controlled area evacuation, (2) a failure to disseminate emergency response information to the public using facilities in the River Bend Station owner controlled area, and (3) a failure to update the emergency plan and procedures after the public was permitted access to facilities in the owner controlled area. The licensee has entered these issues into its corrective action program in CR-RBS-2001-1713 and CR-RBS-2002-0183. This issue was preliminarily determined to have substantial safety significance (Yellow) because it represented a failure to meet a risk-significant emergency preparedness planning standard. UPDATE: On July 31, 2002, a Notice of Violation (EA-02-036) was issued regarding this issue. The violation was as follows: 10 CFR 50.54(q) file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 13 of 19 strates, in part, that a licensee authorized to possess and operate a nuclear power reactor shall follow and maintain in effect emergency plans which meet the standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(7) requires that onsite emergency response plans for nuclear power reactors meet the following standard, which states, in part: "Information is made available to the public on a periodic basis on how they will be notified and what their initial actions should be in an emergency..." Contrary to the above, between 1985 and February 1, 2002, the licensee's emergency plan was not adequate to assure that information was made available to members of the public using River Bend Station's owner conrolled area regarding how members of the public would be notified of an evacuation order and what their initial actions should be in an emergency. Specifically, the licensee had not provided information to members of the public using the West Feliciana Community Development Foundation, the security firing range, the activity center, the outage campground, the Sportsman's Association base camp, and adjacent hunting and fishing areas in the owner controlled area about: (1) the process used to notify the public of an emergency, (2) circumstances under which the public in the licensee's owner controlled area would be directed to assembly and radiological monitoring stations, (3) the predetermined locations of the assembly and radiological monitoring stations, (4) evacuation routes to the predetermined assembly and radiological monitoring stations, and (5) the radiological monitoring and decontamination process. This violation is associated with a White Significance Determination Process finding.
Inspection Report# : 2002005(pdf)
Significance:        Mar 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Reduction in Emergency Plan Effectiveness without Prior NRC Approval This noncited violation is described in a letter to the licensee dated July 31, 2002, regarding the "Final Significance Determination for a White Finding and Notice of Violation." Green. A noncited violation of very low risk significance was identified for failure to comply with the requirements of 10 CFR 50.54(q). Between 1985 and January 2002, the licensee reduced the effectiveness of its emergency plan without Commission approval when it: (1) changed from the use of security vehicles equipped with permanently-mounted public address systems to the use of vehicles without such systems, and relied on portable public address systems stored onsite, (2) canceled emergency plan implementing procedure EIP-2-026, "Evacuation, Personnel Accountability, and Search and Rescue," Revision 11, and (3) permitted several changes in the public's use of the River Bend Station owner controlled area without evaluation of the impact of those changes on the emergency plan. 10 CFR 50.54(q) requires, in part, that each nuclear power plant licensee may make changes to its emergency plans without Commission approval only if the changes do not decrease the effectiveness of the plans and the plans, as changed, continue to meet the standards of 10 CFR 50.47(b) and the requirements of Appendix E of 10 CFR Part 50. The decrease in effectiveness of the emergency plan resulting from the failure to evaluate changes in the station owner controlled area, changes to emergency plan implementing procedures, and changes in emergency notification methods used by security officers, was a performance deficiency. The finding was more than minor because it was associated with one of the Emergency Preparedness cornerstone attributes (Plan Changes) and affected the associated cornerstone objective. Using the Emergency Preparedness Significance Determination Process, the inspector determined the violation had very low risk significance because the violation did not constitute a failure to meet an emergency planning standard as defined by 10 CFR 50.47(b). Because of the very low safety significance and because the licensee included the finding in their corrective action program as Condition Report 2002-0183, this finding is being treated as a noncited violation in accordance with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2002005(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 14 of 19 Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A.
Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:      Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A.
Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Occupational Radiation Safety Significance:      Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:      Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 15 of 19 Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys," requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:        Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inaudible alarm for personal electronic dosimeter used in a high radiation area Technical Specification 5.7.1.b states, in part, that any individual or group of individuals permitted to enter a high radiation area shall be provided with a radiation monitoring device that continuously integrates the radiation dose rate and alarms when a preset integrated dose is received. On October 5, 2001, the licensee identified that an individual working in a high radiation area was unable to hear his electronic dosimeter alarming on the dose accumulated alarm.
Because the individual was unable to respond to the aural alarm, the device was inadequate to fulfill its Technical Specification required function. This violation is being treated as a noncited violation and is in the licensee's corrective action program as CR-RBS-2001-1325. 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, no substantial potential for overexposure, and no impact on the ability to assess dose.
Inspection Report# : 2001007(pdf)
Significance:        Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed. Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Significance:        Jun 23, 2001 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 16 of 19 Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions. Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:      Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:      Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls.
Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:      Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                      Page 17 of 19 Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters. Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection Significance:      Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:      May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 18 of 19 Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation."
The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards. Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability, was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - River Bend 1                                                                    Page 19 of 19 Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/03/2003
 
3Q/2002 Inspection Findings - River Bend 1                                                                        Page 1 of 18 River Bend 1 Initiating Events Significance:      Jun 07, 2001 Identified By: NRC Item Type: FIN Finding Failure to specify or document postmaintenance test requirements.
The inspectors identified that the licensee failed to specify or document postmaintenance test requirements in two main feedwater pump mechanical seal replacement work packages. The failure to specify and document postmaintenance testing for maintenance work activities precluded the ability to evaluate test results to ensure the affected equipment was capable of performing its design function. The inspectors determined that corrective actions for prior postmaintenance testing program deficiencies failed to preclude the recently identified deficiencies. The safety significance of the failure to specify or document postmaintenance test requirements in the two feedwater pump work packages was very low. The issue would not contribute to both the likelihood of an initiating event and the failure of mitigating equipment. Only two of the three main feedwater pumps were affected and only one main feedwater pump is required for mitigation of the reactor trip transient. This finding is documented in the licensee's corrective action program as CR-RBS-2001-0695.
Inspection Report# : 2001003(pdf)
Mitigating Systems Significance:      Aug 15, 2002 Identified By: NRC Item Type: FIN Finding Ineffective corrective actions caused station blackout diesel generator to be unavailable On August 15, 2002, the licensee performed a routine monthly performance test of the station blackout diesel generator. Four minutes into the one-hour run the diesel generator tripped on high coolant temperature. Similar failures of the station blackout diesel generator to run due to high temperature trips had occurred in each of the two previous monthly performance tests on June 21 and July 19, 2002. For each of these failures, the licensee identified an apparent cause for the failure and corrected the problems identified. Following the failure on August 15, 2002, the inspectors determined that the licensee-identified causes for the previous station blackout diesel generator failures were not accurate; therefore, the corrective actions taken were ineffective. The inspectors evaluated the ineffective corrective actions taken to correct two failures of the station blackout diesel generator using inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The inspectors determined that the finding was more than minor in that it affected the operability and availability of a risk-significant mitigating system, i.e., the station blackout diesel generator. The inspectors determined that the failure to maintain the station blackout diesel generator operable was of very low safety significance (Green) because of the low likelihood of a station blackout event occurring, the probability that operators could restore the diesel following an initial failure, and the availability of all other standby electrical systems. This problem identification and resolution issue was entered into the licensee's corrective action program as CR-RBS-2002-0664.
Inspection Report# : 2002003(pdf)
Significance:      May 29, 2002
 
3Q/2002 Inspection Findings - River Bend 1                                                                      Page 2 of 18 Identified By: NRC Item Type: FIN Finding Increased Division I Emergency Diesel Generator jacket cooling water leak rate caused diesel generator to be operable but degraded beyond the licensee's existing evaluation Following maintenance performed on May 9, 2002, to determine the source of a leak from the Division 1 emergency diesel generator jacket cooling water system, the leak rate more than doubled. The licensee's attempt to correct the problem on May 30, 2002, resulted in another increase in the leak rate to the point that makeup to the jacket cooling water system would be required within approximately 2 hours of Division I emergency diesel generator operation during a loss of offsite power. Although, the cause for the increased jacket water leak was repaired on June 4, 2002, the diesel generator remained degraded, but operable. The licensee planned to repair the original leak during the next extended diesel generator maintenance outage. The inspectors determined that the increased leak rate was beyond the licensee's evaluation that concluded that the Division 1 emergency diesel generator was degraded but operable. If left uncorrected, the jacket cooling water leak could have caused the emergency diesel generator to become inoperable and unavailable. The normal source of makeup water would not have been available during a loss of offsite power and the licensee did not develop a written procedure for use of an alternate makeup source until May 30, 2002. Using the significance determination process, the risk significance of the finding was determined to be very low because the emergency diesel generator remained operable, although degraded. This maintenance induced problem was documented in the licensee's corrective action program as Condition Report CR-RBS-2002-0672.
Inspection Report# : 2002002(pdf)
Significance:      May 12, 2002 Identified By: Self Disclosing Item Type: FIN Finding Operator action caused a high reactor water level trip of the running reactor feed pump following a planned scram from 26 percent power Following a planned reactor scram during a plant shutdown, operators failed to take manual control of the feedwater level control system in time to stop an unexpected rise in reactor water level until after the running reactor feed pump tripped on high reactor water level. The licensee determined that the reduction of the reactor pressure control setpoint and subsequent opening of the main turbine bypass valves caused a "swell" in reactor water level which contributed to the higher than expected reactor water level transient. The inspectors determined that the operators did not manually close and isolate one of the two automatic feedwater regulating valves in time to eliminate leakage past the feedwater regulating valve, and failed to reject water from the reactor through the reactor water cleanup system in time to stop the rise in reactor water level to the high level trip of the reactor feed pump. The failure of the operators to manually control reactor water level resulted in the unavailability of a risk-significant reactor feed pump. The inspectors, using the significance determination process, determined that the safety significance of the high reactor water level trip of the running reactor feed pump following a planned reactor scram was very low because the reactor feed pump was restarted from the main control room as soon as reactor water level was lowered and the high reactor water level trip signal was cleared, and other reactor water makeup sources remained available. This human performance error was documented in the licensee's corrective action program as Condition Report CR-RBS-2002-0688.
Inspection Report# : 2002002(pdf)
Significance:      May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding Station Blackout Diesel Generator inoperable due to discharged starting battery The station blackout diesel generator was found to be inoperable by the licensee because its starting battery had been allowed to completely discharge. The station blackout diesel generator had been moved from its normal storage location as a contingency for a planned maintenance outage on several Division I safety-related systems. The inspectors determined that the Division I maintenance outage contingency plan and the weekly work schedule did not plan for the return of the station blackout diesel generator to its normal storage location to re-energize its battery charger. The licensee determined that this is a repeat of a similar event of April 4, 1998, documented in Condition Report CR-RBS-1998-0384. The failure to maintain its starting battery charged caused the risk significant station blackout diesel
 
3Q/2002 Inspection Findings - River Bend 1                                                                      Page 3 of 18 generator to be inoperable and unavailable. The inspectors, using the significance determination process, determined that the safety significance of the unavailability of the station blackout diesel generator was very low because the length of time the diesel generator was unavailable was less than 24 hours and all other electrical systems were available during that time. This human performance error was documented is the licensee's corrective action program as Condition Report CR-RBS-2002-0664.
Inspection Report# : 2002002(pdf)
Significance:      Dec 07, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to control test equipment when it was considered unreliable during a Technical Specification surveillance and failure to evaluate initial out of tolerance test data The licensee did not control measuring and test equipment when it was considered to be unreliable during a reactor core isolation cooling system surveillance test and did not evaluate the initial out of tolerance data to ensure the original test results were not valid. Specifically, measuring and test equipment originally indicated a suppression pool level instrument failed a Technical Specification surveillance test, so the measuring and test equipment was considered to be unreliable. Another piece of measuring and test equipment was then used and the suppression pool level instrument passed the surveillance test. The inspectors identified that maintenance personnel did not control the original measuring and test equipment for subsequent calibration checks and did not notify operations personnel to evaluate the original out of specification data to ensure the original test results were not valid, as required by plant procedures for suspect measuring and test equipment. The inspectors determined that the safety significance of failing to control measuring and test equipment and then to evaluate the original test data was very low since it did not represent an actual loss of the reactor core isolation cooling system or suppression pool reliability. The failure to control unreliable measuring and test equipment and to evaluate test results provided by such equipment is a noncited violation of Technical Specification 5.4.1a. This human performance error is documented in the licensee's corrective action program as CR-RBS-2001-1650.
Inspection Report# : 2001004(pdf)
Significance:      Aug 01, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain fire barrier requirements described in the plant fire hazards analysis.
The licensee did not maintain a 3-hour rated fire barrier as described in the plant fire hazards analysis. Specifically, the inspectors identified a penetration into a 3-hour rated floor barrier in the standby cooling tower that had not been sealed. The inspectors determined that the safety significance of the degraded fire barrier was very low since it did not separate redundant safe shutdown equipment. The failure to maintain a 3-hour rated fire barrier as described in the Fire Hazards Analysis is a noncited violation of Attachment 4 to Facility Operating License NPF-47. This violation is documented in the licensee's corrective action program as CR-RBS-2001-0898.
Inspection Report# : 2001003(pdf)
Significance:      Jul 11, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies with the conduct of maintenance risk assessments.
The inspectors identified deficiencies with the conduct of maintenance risk assessments. Specifically, inadequate risk assessments were identified, a plant component was not identified by the licensee to be in the quantitative risk assessment tool, an opportunity to identify an error in the risk assessment tool was missed, and corrective actions taken for prior inadequate risk assessments failed to preclude the recently identified deficiencies. The inspectors determined that the safety significance of the maintenance risk assessment deficiencies was very low in that there was no actual loss of safety function and that the difference between the actual plant risk and the licensee determined risk was small enough such that significant risk management actions would not have been required. This finding is documented in the
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 4 of 18 licensee's corrective action program as CR-RBS-2001-0674.
Inspection Report# : 2001003(pdf)
Significance:      Mar 31, 2001 Identified By: NRC Item Type: FIN Finding Deficiencies With Implementation of Cold Weather Requirements The inspectors identified deficiencies with implementation of cold weather requirements. Specifically, no attempt was made to provide heating to the water treatment room, several room heater switch settings were not in accordance with the respective loop calibration report, and repetitive maintenance tasks did not exist to ensure that room heaters in the fire pump building, the normal switchgear room, or the auxiliary control room were functioning properly. The inspectors determined that the safety significance of not implementing cold weather requirements for reduced room temperatures was very low in that actual temperatures in these areas during the time the condition existed did not go low enough to affect the qualification of the equipment located in these areas.
Inspection Report# : 2000016(pdf)
Significance:      Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain the Required Inventory of Alternate Standby Liquid Control Chemicals The licensee failed to maintain the required inventory of chemicals onsite to support operation of the alternate standby liquid control system. Specifically, the licensee failed to maintain 2500 pounds of sodium borate and 2500 pounds of boric acid onsite for alternate standby liquid control injection as required by the emergency operating procedure. The inspectors determined that the safety significance of not maintaining alternate standby liquid control chemicals available was very low in that the standby liquid control system was only determined to be unavailable for a maximum of 12 days over the year during tank sparging evolutions. The failure to maintain adequate chemical inventory in the main warehouse/storeroom for alternate standby liquid control injection is a noncited violation of Technical Specification 5.4.1.a. This violation is in the licensee's corrective action program as Condition Report 2000-1680.
Inspection Report# : 2000016(pdf)
Significance:      Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Complete Annual Walkdowns of Emergency Operating Procedure Enclosures The licensee did not complete annual walkdowns of emergency operating procedure enclosures between November 1996 and October 2000. The inspectors determined that the safety significance of not completing annual walkdowns of emergency operating procedure enclosures was very low in that, other than missing alternate standby liquid control chemicals, no significant equipment issues were identified when the enclosures were walked down. Additionally, no actual plant problems occurred which would have required implementation of these enclosures. The failure to perform yearly walkdowns of each emergency operating procedure enclosure, as required by Procedure OSP-0009, is a noncited violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is in the licensee's corrective action program as Condition Report 2000-1723.
Inspection Report# : 2000016(pdf)
Significance:      Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain A Fire Barrier Between Two Fire Areas Which Contain Redundant Safe Shutdown Equipment
 
3Q/2002 Inspection Findings - River Bend 1                                                                  Page 5 of 18 The licensee did not maintain a 3-hour rated fire barrier between two fire areas which contained redundant safe shutdown equipment. Specifically, the inspectors identified an 11.5-inch deep hole in a 12-inch concrete fire barrier between the D-Tunnel and the D-Tunnel cable chase fire areas. The inspectors determined that the safety significance of the degraded fire barrier was very low due to the remaining mitigating detection and suppression systems, the fire brigade response, and the low initiating frequency. The failure to maintain a 3-hour rated fire barrier between Fire Areas AB-7 and -18, is a noncited violation of Attachment 4 to Facility Operating License 50-458. This violation is in the licensee's corrective action program as Condition Report 2000-1944.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure To Monitor the Performance of a Standby Service Water Component Against Established Goals To Ensure It Was Capable of Performing Its Maintenance Rule Function The licensee did not monitor the performance of standby service water station blackout Valve SWP-AOV599 against established goals in a manner sufficient to assure the valve was capable of supplying cooling water to the Division III emergency diesel generator during a station blackout event. The inspectors determined that the safety significance of the failure to monitor the station blackout valve was very low due to the high likelihood of success of operator recovery actions. The failure to monitor the performance of Valve SWP-AOV599 is a noncited violation of 10 CFR 50.65(a)(1).
This violation (EA-01-090) is in the licensee's corrective action program as Condition Report 1999-0263.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified Failure To Implement Corrective Actions For a Condition Adverse To Quality Criterion XVI of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. The licensee determined that they failed to implement effective corrective actions to correct a condition adverse to quality involving the performance of loss of offsite power logic system functional testing. Consequently, during Refueling Outage 8, the licensee did not adequately perform Technical Specification Surveillance Requirement 3.3.8.1.4, which required that a logic system functional test of the loss of offsite power instrumentation be completed every 18 months. The issue is described in the licensee's corrective action program reference Condition Report 2000-1813 and Licensee Event Report 50-458/0015.
Inspection Report# : 2000016(pdf)
Significance: SL-IV Dec 23, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to obtain Commission approval for a change to the USAR The licensee made a change to the fuel handling accident dose reported in the Updated Safety Analysis Report for the exclusion area boundary thyroid that represented an increase in consequences without obtaining prior Commission approval as required by 10 CFR 50.59. This violation of 10 CFR 50.59(b)(1) identified above is categorized at Severity Level IV and is being treated as a noncited violation, consistent with Section VI.A1 of the NRC Enforcement Policy.
This violation (50-458/0015-01) (EA-00-267-1) was entered into the licensee's corrective action program as Condition Report 2000-2050 (Section 1R02.b). This finding was of very low safety significance because previous and subsequent doses for the fuel handling accident exclusion area boundary thyroid were greater than the value implemented by this change.
Inspection Report# : 2000015(pdf)
Significance:        Nov 11, 2000 Identified By: NRC
 
3Q/2002 Inspection Findings - River Bend 1                                                                      Page 6 of 18 Item Type: FIN Finding Poor operability assessment of station blackout valve Engineering personnel did not properly assess the significance of system air leakage on the ability to maintain station blackout Valve SWP-AOV599 open for the 12-hour duration specified in the probabilistic safety assessment.
Specifically, engineering personnel only considered the minimum air pressure necessary to open the valve and did not determine the minimum air pressure needed to maintain the valve in the open position. The poor engineering review of air leakage on station blackout Valve SWP-AOV599 was of very low safety significance in that subsequent air drop testing of the system and engineering analysis demonstrated that the valve would have remained open for the 12-hour duration specified in the probabilistic safety analysis.
Inspection Report# : 2000014(pdf)
Significance:        Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately assess and conduct fire drills The licensee did not adequately assess or conduct fire drills. During the October 11, 2000, fire brigade drill, the licensee failed to identify and assess several deficiencies. For example, brigade members incorrectly donned protective clothing, the brigade leader did not establish communications between the control room and scene, there was no simulated demonstration of the ability to pressurize a hose or use a hose nozzle, two brigade members did not actively participate in the simulated extinguishing of the fire, and objective criteria were not developed to evaluate the fire brigade's performance. Additionally, the licensee performed unannounced drills within 4 weeks of each other and did not use members of the management staff responsible for plant safety and fire protection to critique unannounced drills.
The failure to adequately assess the effectiveness of the fire brigade and to adequately conduct fire brigade drills was a violation of Attachment 4 to Facility Operating License 50-458. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1848. The inspectors determined that the safety significance of the fire brigade training issues and fire brigade performance was very low in that plant fire barriers and automatic suppression capability were maintained in accordance with the fire protection program.
Inspection Report# : 2000014(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement the Maintenance Rule The licensee failed to identify that a failure of the Division I hydrogen igniter to start in 1999 was a maintenance preventable functional failure. Consequently, when the same failure occurred in 2000, a repeat maintenance preventable functional failure was not identified. As a result, the hydrogen igniter system was not assessed as required for inclusion under the licensee's maintenance rule provisions of 10 CFR 50.65(a)(1). This was identified as a violation of 10 CFR 50.65(a)(1) and additionally of 10 CFR 50.65(a)(2), since the performance monitoring provisions of this section were not properly accomplished. This violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (EA-00-242) (50-458/0017-01) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1762.
Inspection Report# : 2000017(pdf)
Significance:        Oct 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Improper minimum voltage assumed in design calculations The specified minimum voltage on the ac buses used to calculate equipment operability was based on an assumption of 95 percent nominal voltage at the Fancy Point substation in lieu of the more limiting technical specification allowable value for the degraded grid voltage relays on the 4.16 kV buses. The technical specification bases stated that these
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 7 of 18 relays were set high enough to ensure that sufficient power was available to the required equipment. However, design calculations did not exist to support this statement. The non-conservative voltage assumption resulted in overestimating the minimum voltage available for motor-operated valves and other loads on the safety-related 480 Vac buses. This discrepancy was identified as a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This violation (50-458/0017-02) was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1764.
Inspection Report# : 2000017(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: FIN Finding Review of functional failure criteria for inoperable but available structures, systems, and components The inspectors determined that engineering personnel did not properly characterize a maintenance activity associated with Valve E12-F067, which unexpectedly isolated residual heat removal Train C, as a maintenance preventable functional failure. The licensee's maintenance rule determination incorrectly assumed that a functional failure could not occur if the system was considered Technical Specification inoperable. This closes Unresolved Item 50-458/0011-04.
The safety significance of this issue was very low because the additional maintenance preventable functional failure did not result in the residual heat removal system exceeding a maintenance rule performance monitoring criteria of less than or equal to one maintenance preventable functional failure in an 18-month period. Additionally, two redundant trains of low pressure coolant injection remained available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly install scaffolding The inspectors determined that scaffold components were installed in contact with permanent plant equipment without prior engineering approval. During tours of the plant between July 10 and September 7, 2000, the inspectors identified incorrectly installed scaffolding which contacted systems involving: control air, standby gas treatment, the main plant exhaust stack, and 480 volt switchgear. Additionally, scaffolding was identified which could have affected the operation of an auxiliary building ventilation system damper. The failure to properly install plant scaffolding as required by plant procedures was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Reports 2000-1350, 2000-1577, 2000-1584, and 2000-1657. The inspectors determined that the safety significance of the improperly installed scaffolding was very low because redundant components not affected by scaffolding were available.
Inspection Report# : 2000013(pdf)
Significance:        Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions for technical deficiencies associated with procedures The inspectors determined that the licensee did not implement corrective actions for identified safety-related operations procedural technical deficiencies. Between April 3, 1995, and June 14, 2000, operations personnel did not implement corrective actions to revise eight operating procedures following the licensee's identification of technical deficiencies with the documents. The failure to properly identify and resolve technical deficiencies in procedures was a cross-cutting issue which was representative of a programmatic problem which had the potential to impact safety in that:
operations personnel were not familiar with the procedure revision process, supervisory or peer reviews were typically not completed for procedure action requests, the operations procedure group was not aware of the content of the procedure revision backlog, quality assurance audits of procedure controls did not assess the content of the procedure backlog, periodic reviews of most operating procedures were not performed, and technical deficiencies with operations
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 8 of 18 procedures remained uncorrected for several years. The failure to implement corrective actions for conditions adverse to quality was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1442. The inspectors determined that the technical deficiencies associated with the procedures were of very low safety significance because, although the deficiencies could have caused some confusion and delay, trained operators would likely have been able to recognize the deficiencies and take the appropriate actions.
Inspection Report# : 2000013(pdf)
Significance:      Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Four examples of the failure to have adequate procedures or follow procedures The inspectors identified four examples of a failure to have adequate procedures or follow procedures. Specifically, offgas system procedures did not provide instructions which were appropriate to the circumstances in that no limitations on air purge flow rates were specified when operating offgas air purge Valves N64-F004A and -B.
Consequently, on August 21, 2000, operations personnel fully opened an air purge valve which resulted in a backpressure on the steam jet air ejectors and subsequent insertion of a manual reactor scram due to lowering main condenser vacuum. Additionally, during reviews of operability evaluations between August 31 and September 11, 2000, the inspectors identified inadequate operability evaluations involving: an inverter, 480 volt breakers, and standby cooling tower switchgear room ventilation. The failure to provide offgas system procedures with instructions appropriate to the circumstances and the failure to adequately perform operability evaluations as required by plant procedures was a violation of Criterion V of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action system as Condition Reports 2000-1506, 2000-1553, 2000-1572, and 2000-1583. The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions (Sections 1R14 and 1R15). The inspectors determined that the safety significance of the inadequate procedure and loss of main condenser vacuum with manual reactor scram event was very low. The reactor trip was uncomplicated and the main condenser remained in service throughout the duration of the scram recovery actions. Additionally, all equipment and operator responses following the event were appropriate. The safety significance of the inadequate operability evaluations was also very low in that subsequent operability evaluations determined that the affected components would have performed their intended safety functions.
Inspection Report# : 2000013(pdf)
Significance:      Aug 05, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to locked valves The inspectors determined that the licensee did not implement adequate corrective actions in response to noncited violation 50-458/9913-01, to ensure manual valves in the main flow path of safety related systems were locked.
Consequently, the inspectors identified approximately 70 manual valves which were not locked as required by plant procedures and the Updated Safety Analysis Report. The failure to implement corrective actions was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This issue was entered into the licensee's corrective action system as Condition Reports 1999-1557 and 2000-1405. The safety significance of this issue was very low because the unlocked manual valves were in the correct position for plant operation. Therefore, the safety function of the associated systems was not affected.
Inspection Report# : 2000011(pdf)
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 9 of 18 Significance:        Aug 05, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to isolate drywell purge isolation valve penetrations The licensee failed to adequately perform surveillance testing to verify that the drywell purge isolation valves were sealed closed. Consequently, two drywell penetrations were inoperable during MODE 1 operations. The failure to seal closed the affected drywell purge isolation valve penetrations by isolating their motive air was a violation of Technical Specification 3.6.5.3. The circumstances involving this issue were discussed in Licensee Event Report 50-458/0009.
This issue was entered into the licensee's corrective action system as Condition Report 2000-1139. The safety significance of this issue was very low because the drywell purge isolation valves were administratively controlled by tags in the main control room and a caution note in plant procedures specified that drywell purge was not to be operated while in MODE 1, 2, or 3. Therefore, the inspectors determined that the drywell purge valves should have remained closed during accident conditions.
Inspection Report# : 2000011(pdf)
Significance: SL-IV Aug 05, 2000 Identified By: NRC Item Type: VIO Violation Failure to complete monthly inspections of portable fire extinguishers The inspectors determined that fire protection personnel did not implement corrective actions to restore compliance in response to a minor violation identified on April 10, 2000, which involved the failure to complete inspections of portable fire extinguishers located in high radiation areas. During tours of the auxiliary building on June 25, 2000, the inspectors again determined that fire protection personnel were not completing inspections of portable fire extinguishers located in high radiation areas. The failure to perform inspections of fire extinguishers was a Severity Level IV violation of License Condition 1 of Attachment 4 to Facility Operating License No. NPF-47. This issue was entered into the licensee's corrective action system as Condition Report 2000-0969. Fire protection personnel failed to implement corrective actions to restore compliance within a reasonable period of time. The safety significance of this issue was very low because redundant methods of automatic and manual fire suppression were available.
Inspection Report# : 2000016(pdf)
Inspection Report# : 2000011(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: FIN Finding Unnecessary Disabling of Suppression Pool Cooling Function The inspectors determined that engineering personnel provided inaccurate information to operations personnel on the functional capability of the residual heat removal heat exchanger bypass valve following the inspectors' discovery that the antirotation device had fallen off. Consequently, operations personnel took conservative action to disable the suppression pool cooling function of residual heat removal Train A for approximately 36 hours. Disabling the residual heat removal Train A suppression pool cooling function had a small impact on safety and affected the safety function of a train of a mitigating system. This issue was of very low risk significance because redundant methods of suppression pool cooling remained operable and unavailability time was less than that allowed by the Technical Specifications.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to specify postmaintenance requirements in maintenance packages The inspectors determined that planning personnel failed to identify required postmaintenance testing requirements in four maintenance packages. The failure to identify the appropriate postmaintenance testing requirements as required by
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 10 of 18 planning procedures was considered a violation of Technical Specification 5.4.1.a. This issue was entered into the licensee's corrective action program as Condition Reports 2000-1010 and 2000-1199. The risk significance of this issue was very low because in-process maintenance activities provided assurance that the affected components were functionally capable.
Inspection Report# : 2000010(pdf)
Significance:        Jun 24, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform functional test of containment unit coolersupply valve The failure to perform functional testing of standby service water supply Valve SWP-MOV502B following breaker maintenance resulted in the Division II primary containment unit cooler being inoperable while the facility was in MODE 1 between February 9 and March 4, 2000. The failure to restore the Division II containment unit cooler within 7 days with the facility in MODE 1 was considered a violation of Technical Specification 3.6.1.7. The circumstances involving this issue were discussed in Licensee Event Report 50-458/00-05. This issue was entered into the licensee's corrective action program as Condition Report 2000-0736. The inspectors and a senior reactor analyst used the significance determination process to evaluate the risk significance of this issue. The most limiting initiating event was an anticipated transient without scram. The risk significance for this event was very low because one containment unit cooler and two residual heat removal trains in the suppression pool cooling mode were available for mitigation.
Inspection Report# : 2000010(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Control Room Configuration Not Consistent With Simulator The inspectors identified three simulator fidelity issues during a walkdown of selected panels in the simulator which involved an out of service reboiler vent valve, an out of service suppression pool temperature indication, and an elevated containment temperature indication. Additionally, the licensee identified four deficiencies during a subsequent audit which involved a feedwater heater controller, a deenergized regenerative evaporator supply shut-off valve, an average power range monitor, and suppression pool cooling Pump 1B. The risk significance of this issue was very low because the deficiencies would not have significantly impacted the effectiveness of simulator training.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Inadvertent Engineered Safety Features Isolation During the restoration of the reactor core isolation cooling system following the steam supply pressure low channel functional test, an inadvertent engineered safety features actuation resulted in the isolation of the reactor core isolation cooling system. The subsequent investigation of the event by engineering personnel determined that instrument and controls personnel inadvertently contacted an adjacent terminal which caused an engineered safety features actuation of the reactor core isolation cooling system. The risk significance of the issue was very low because additional injection systems were operable.
Inspection Report# : 2000009(pdf)
Significance:        May 06, 2000 Identified By: NRC Item Type: FIN Finding Residual heat removal function not included in the Maintenance Rule scope
 
3Q/2002 Inspection Findings - River Bend 1                                                                  Page 11 of 18 The function of the residual heat removal minimum flow valves, as described in the bases for Technical Specification 3.3.5.1, "Emergency Core Cooling System Instrumentation," was not included in the list of functions included in the maintenance rule scope for the residual heat removal system. Consequently, maintenance rule functional failures associated with residual heat removal minimum flow Valve E12-F064A opening when aligning the residual heat removal system to the shut down cooling mode of operation were not identified by engineering personnel. The risk significance of this issue was very low because the improper characterization of the failure of Valve E12-F064A did not significantly impact implementation of the maintenance rule for the residual heat removal system.
Inspection Report# : 2000009(pdf)
Significance:      May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement corrective actions to prevent recurrence of inadvertent opening of residual heat removal minimum flow valve The licensee did not implement corrective actions following a previous occurrence to preclude opening of residual heat removal minimum flow Valve E12-F064A and subsequent loss of approximately 50 gallons of reactor vessel inventory while aligning the residual heat removal system to the shutdown cooling mode of operation. The risk significance of this issue was very low because redundant methods of inventory injection were either operating or available. This item was entered in the licensee's corrective action program as Condition Report 2000-0947.
Inspection Report# : 2000009(pdf)
Significance:      May 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Three examples of a failure to follow procedures involving debris in the drywell, unqualified coatings in the drywell, and scram time testing Maintenance and engineering personnel did not adequately perform a zone inspection of the drywell as required by Maintenance Action Item 329427, "Drywell Zone Inspection - All Levels." Specifically, the inspectors identified a significant amount of debris during a drywell closeout inspection which had not been identified during the licensee's zone inspection or during a management closeout tour. Maintenance and engineering personnel did not adequately perform a coatings inspection of the drywell as required by Maintenance Action Item 333068, "Drywell Coating Inspection." Specifically, the inspectors identified 400 to 500 square feet of degraded coatings during a drywell closeout inspection which had not been identified during the licensee's coatings inspection or during a management closeout tour. The risk significance of the drywell issues was very low because the emergency core cooling system suction strainers would not have been adversely affected. Operations and engineering personnel did not complete a control rod drop accident analysis as required by Procedure STP-500-0705, "Rod Sequence Verification When Rod Pattern Control System Is Bypassed." Specifically, operations personnel withdrew Control Rod 44-13 beyond the banked position withdrawal sequence restraints without having completed a control rod drop accident analysis. The risk significance of this issue was low because the licensee subsequently determined that the plant remained within the boundaries of the control rod drop accident analysis. These items were entered in the licensee's corrective action program as Condition Reports 2000-0911, 2000-0904, and 2000-0941.
Inspection Report# : 2000009(pdf)
Barrier Integrity Significance:      May 06, 2000 Identified By: NRC Item Type: FIN Finding
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 12 of 18 Incomplete Maintenance Rule Determination The licensee's maintenance rule functional failure review of the failure of Valve E51-F076 to close only considered the effect on reactor core isolation cooling system operation and did not evaluate the effect on the containment isolation function. The risk significance of this issue was very low because the improper characterization of the failure of Valve E51-F076 did not significantly impact implementation of the maintenance rule for the reactor core isolation cooling system.
Inspection Report# : 2000009(pdf)
Emergency Preparedness Significance:        Mar 20, 2002 Identified By: NRC Item Type: VIO Violation Failure to develop a range of protective actions, disseminate emergency response information, and maintain the emergency plan for members of the public located within the owner controlled area The inspector identified one preliminary finding involving the failure to develop a range of protective actions, disseminate emergency response information, and maintain the emergency plan in accordance with the requirements of 10 CFR 50.54(q), planning standards §50.47(b)(10) and (7), and 10 CFR Part 50, Appendix E, section IV(G) pertaining to members of the public located in the owner controlled area. Three apparent violations are associated with the finding. The issues involved: (1) a failure to establish effective means or provisions for warning, advising, evacuating, and monitoring members of the public during an owner controlled area evacuation, (2) a failure to disseminate emergency response information to the public using facilities in the River Bend Station owner controlled area, and (3) a failure to update the emergency plan and procedures after the public was permitted access to facilities in the owner controlled area. The licensee has entered these issues into its corrective action program in CR-RBS-2001-1713 and CR-RBS-2002-0183. This issue was preliminarily determined to have substantial safety significance (Yellow) because it represented a failure to meet a risk-significant emergency preparedness planning standard. UPDATE: On July 31, 2002, a Notice of Violation (EA-02-036) was issued regarding this issue. The violation was as follows: 10 CFR 50.54(q) strates, in part, that a licensee authorized to possess and operate a nuclear power reactor shall follow and maintain in effect emergency plans which meet the standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(7) requires that onsite emergency response plans for nuclear power reactors meet the following standard, which states, in part: "Information is made available to the public on a periodic basis on how they will be notified and what their initial actions should be in an emergency..." Contrary to the above, between 1985 and February 1, 2002, the licensee's emergency plan was not adequate to assure that information was made available to members of the public using River Bend Station's owner conrolled area regarding how members of the public would be notified of an evacuation order and what their initial actions should be in an emergency. Specifically, the licensee had not provided information to members of the public using the West Feliciana Community Development Foundation, the security firing range, the activity center, the outage campground, the Sportsman's Association base camp, and adjacent hunting and fishing areas in the owner controlled area about: (1) the process used to notify the public of an emergency, (2) circumstances under which the public in the licensee's owner controlled area would be directed to assembly and radiological monitoring stations, (3) the predetermined locations of the assembly and radiological monitoring stations, (4) evacuation routes to the predetermined assembly and radiological monitoring stations, and (5) the radiological monitoring and decontamination process. This violation is associated with a White Significance Determination Process finding.
Inspection Report# : 2002005(pdf)
Significance:        Mar 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Reduction in Emergency Plan Effectiveness without Prior NRC Approval This noncited violation is described in a letter to the licensee dated July 31, 2002, regarding the "Final Significance Determination for a White Finding and Notice of Violation." Green. A noncited violation of very low risk significance
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 13 of 18 was identified for failure to comply with the requirements of 10 CFR 50.54(q). Between 1985 and January 2002, the licensee reduced the effectiveness of its emergency plan without Commission approval when it: (1) changed from the use of security vehicles equipped with permanently-mounted public address systems to the use of vehicles without such systems, and relied on portable public address systems stored onsite, (2) canceled emergency plan implementing procedure EIP-2-026, "Evacuation, Personnel Accountability, and Search and Rescue," Revision 11, and (3) permitted several changes in the public's use of the River Bend Station owner controlled area without evaluation of the impact of those changes on the emergency plan. 10 CFR 50.54(q) requires, in part, that each nuclear power plant licensee may make changes to its emergency plans without Commission approval only if the changes do not decrease the effectiveness of the plans and the plans, as changed, continue to meet the standards of 10 CFR 50.47(b) and the requirements of Appendix E of 10 CFR Part 50. The decrease in effectiveness of the emergency plan resulting from the failure to evaluate changes in the station owner controlled area, changes to emergency plan implementing procedures, and changes in emergency notification methods used by security officers, was a performance deficiency. The finding was more than minor because it was associated with one of the Emergency Preparedness cornerstone attributes (Plan Changes) and affected the associated cornerstone objective. Using the Emergency Preparedness Significance Determination Process, the inspector determined the violation had very low risk significance because the violation did not constitute a failure to meet an emergency planning standard as defined by 10 CFR 50.47(b). Because of the very low safety significance and because the licensee included the finding in their corrective action program as Condition Report 2002-0183, this finding is being treated as a noncited violation in accordance with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2002005(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to flow test self-contained breathing apparatus regulators Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A.
Section 4.8 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus regulators to be flow tested in accordance with the manufacture's recommendations every 2 years. On April 30, 2001, the licensee identified 48 self contained breathing apparatus regulators that were past due for their 2-year flow test. This event is documented in the licensee's corrective action program as CR-RBS-2001-0551. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
Significance:        Dec 20, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform hydrostatic testing of self contained breathing apparatus cylinders Technical Specification 5.4.1 requires the implementation of procedures listed in Regulatory Guide 1.33, Appendix A.
Section 4.11 of Procedure RPP-022, "Respiratory Protection Equipment Cleaning, Inspection, and Repair," requires self-contained breathing apparatus cylinders be inspected and undergo hydrostatic testing every 3 years by a Department of Transportation approved test vendor. On December 12, 2001, the licensee identified 48 self-contained breathing apparatus cylinders that were in use and had not been hydrostatically tested within the last 3 years. This event is documented in the licensee's corrective action program as CR-RBS-2001-1666. This violation is being treated as a noncited violation. The safety significance of this violation was determined to be very low by the Emergency Preparedness Safety Significance Determination Process because there was no failure to meet an emergency planning standard or risk significant planning standard.
Inspection Report# : 2001004(pdf)
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 14 of 18 Occupational Radiation Safety Significance:      Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to reduce dose margin 10 CFR 20.1201(f) requires licensees to reduce the dose that an individual may be allowed to receive in the current year by the amount of dose received while employed by any other person. On July 25, 2001, the licensee identified that an employee, who supported Grand Gulf Station during their refueling outage and received approximately 1,100 millirem, returned to the site and entered the controlled access area to perform work without having his exposure margin reduced. This event is documented in the licensee's corrective action program as CR-RBS-2001-0860. 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# : 2001004(pdf)
Significance:      Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 violation for failure to follow procedure Technical Specification 5.4.1 requires written procedures to perform radiological surveys. Station Procedure RPP-006, "Radiological Surveys," requires that a survey including dose rates and contamination levels be conducted prior to allowing workers to access radiologically restricted areas that are not surveyed on a routine basis. On October 1, 2001, the licensee identified that workers entered the reactor water cleanup pump room, a locked high radiation area that is not routinely surveyed, without performing a current survey. This event is documented in the licensee's corrective action program as CR-RBS-2001-1264. 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# : 2001004(pdf)
Significance:      Oct 05, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Inaudible alarm for personal electronic dosimeter used in a high radiation area Technical Specification 5.7.1.b states, in part, that any individual or group of individuals permitted to enter a high radiation area shall be provided with a radiation monitoring device that continuously integrates the radiation dose rate and alarms when a preset integrated dose is received. On October 5, 2001, the licensee identified that an individual working in a high radiation area was unable to hear his electronic dosimeter alarming on the dose accumulated alarm.
Because the individual was unable to respond to the aural alarm, the device was inadequate to fulfill its Technical Specification required function. This violation is being treated as a noncited violation and is in the licensee's corrective action program as CR-RBS-2001-1325. 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, no substantial potential for overexposure, and no impact on the ability to assess dose.
Inspection Report# : 2001007(pdf)
Significance:      Sep 28, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 15 of 18 Failure to review work and identify dose saving measures.
The inspector identified a noncited violation of very low safety significance because the licensee's work control process failed to ensure that all work activities were reviewed to identify opportunities to reduce radiation doses. The failure resulted from the lack of an implementing procedure that required the review of temporary electrical power installations to take into account factors for minimizing radiation exposure to workmen, in violation of Technical Specification 5.4.1. A total of 94 temporary power installations were scheduled for the outage but had not been reviewed. Three installations had been completed before the identification of the problem. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee placed this item in its corrective action program as CR-RBS-2001-1149. The failure to implement dose saving measures had a credible impact on safety. The occurrences involved workers' unplanned, unintended doses that resulted from actions that were contrary to licensee procedures and Technical Specifications. However, the safety significance was determined to be very low because there was no exposure in excess of regulatory limits or significant potential for exposure in excess of regulatory limits.
Inspection Report# : 2001003(pdf)
Significance:        Jun 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to keep radiation workers informed of radiological conditions The inspectors identified a noncited violation of 10 CFR 19.12(a) for failure to keep radiation workers informed of radiological conditions. Specifically, personnel did not receive a radiological hazards briefing prior to a high radiation area entry as required by NRC regulations. This finding was greater than minor and had a credible impact on safety because of the potential for unintended and unplanned dose resulting from actual radiological conditions. The inspectors determined that this failure to brief radiation workers prior to entry into a high radiation area was of very low safety significance by the Occupational Radiation Safety Significance Determination Process since it was not an as low as reasonably achievable issue, the ability to assess dose was not compromised, and there was no actual or substantial potential exposure in excess of 10 CFR Part 20 dose limits. The safety significance of the condition was further mitigated by the conservative setpoints on the alarming dosimetry worn by the personnel during the entry.
Inspection Report# : 2001002(pdf)
Public Radiation Safety Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Survey LIcensed Radioactive Materials 10 CFR 20.1501(a) states, in part, each licensee shall make or cause to be made, surveys that are reasonable under the circumstances to evaluate radiation levels, concentrations or quantities of radioactive material, and potential radiological hazards. On October 9, 2000, the licensee identified three examples of detectable licensed radioactivity that was unconditionally released from the controlled access area, as described in the licensee's corrective action program, reference Condition Report 2000-1788.
Inspection Report# : 2000016(pdf)
Significance:        Mar 31, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure To Control Liquid Effluent Release Rates Below the Value Specified On Two Discharge Permits Technical Specification 5.5.1.b requires that the offsite dose calculation manual contain radioactive effluent controls.
 
3Q/2002 Inspection Findings - River Bend 1                                                                      Page 16 of 18 Offsite Dose Calculation Manual 7.2.2.1 states, in part, that release rates shall be administratively controlled to maintain the fraction of 10 CFR Part 20 limits less than or equal to 0.3. Station Procedure SOP-0113, step 5.6.20, requires that the LWS-FIC197 (liquid effluent discharge flow) setpoint is adjusted to the desired flow rate not to exceed the value specified on the liquid radwaste discharge permit. On February 26, 2000 (Permit 2000027) and April 15, 2000 (Permit 2000080), liquid discharges were made which exceeded the maximum allowable release rate, as described in the licensee's corrective action program (Condition Reports 2000-0403 and -0982).
Inspection Report# : 2000016(pdf)
Significance:      Oct 04, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify the radioactive waste in two shipments The inspector identified that the licensee did not properly classify the radioactive waste in two shipments. Radioactive Waste Shipments 00-058 and 00-059 contained sock type mechanical filters; however, there was no 10 CFR Part 61 waste stream analysis for any mechanical filters. Instead, the licensee utilized a bead resin waste stream analysis to classify the shipments. The licensee had not confirmed, through sampling and analysis, that these two waste streams were similar. Because the licensee had not sampled and analyzed the sock type mechanical filter waste stream, it did not provide reasonable assurance that the indirect method of identifying radionuclides was valid. Therefore, the radioactive waste in Radioactive Material Shipments 00-058 and 00-059 were not properly classified in accordance with 10 CFR 61.55 and were two examples of a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This issue was entered in the licensee's corrective action program as Condition Report 2000-1463. The inspectors determined that the improper classification of radioactive material shipments was of very low safety significance because the shipments were not underclassified.
Inspection Report# : 2000013(pdf)
Physical Protection Significance:      Jun 23, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Significance:      May 09, 2001 Identified By: NRC Item Type: FIN Finding Failure to prevent a simulated adversay from gaining access to a vital area During an Operational Safeguards Response Evaluation conducted on June 19-23, 2000, a vulnerability in the licensee's protective strategy was identified that resulted in the simulated loss of part of a target set. Further details (safeguards information) are available in NRC Inspection Report 50-458/2000-12. The issue was entered into the licensee's
 
3Q/2002 Inspection Findings - River Bend 1                                                                  Page 17 of 18 corrective action program as Condition Report CR-RBS-2000-1302. The safety significance of this finding was determined to be very low by the Physical Protection Significance Determination Process because it was not repeatable or predictable. The issue was more than minor because the potential loss of a target set represents a credible impact on safety and impacts a key performance attribute of the physical protection cornerstone.
Inspection Report# : 2001002(pdf)
Miscellaneous Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Overall, an effective corrective action program was in place The licensee was effective at identifying problems and putting them into the corrective action program. However, the licensee's corrective action program procedures did not require an additional review of reportability when an operability determination was subsequently modified. In several instances documentation for past operability and reportability decisions was lacking. However, no instances were identified in which the licensee failed to make a required report. There were instances in which the licensee conducted reviews and evaluations as a part of their corrective actions that were related to events or conditions, but did not document these activities. The licensee implemented corrective actions, when specified, in a timely manner. The licensee performed effective audits and self-assessments. During interviews conducted during this inspection, the site staff expressed open willingness to input safety issues into the problem identification and resolution program.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 13, 2001 Identified By: NRC Item Type: FIN Finding Weak reportability evaluation and past operability assessment performance In several instances, licensee records lacked evidence that evaluations for past-operability assessments were performed when equipment or systems failed routine surveillance tests. Further, in some instances, the licensee determined reportability before relevant evaluations were completed. Finally, as a matter of routine, the licensee did not re-assess the reportability of an event or condition following a revision to an operability determination subsequent to the initial reportability determination. However, no instances were identified in which the licensee failed to make a required report. The NRC evaluated the issue using the significance and documentation determination process of NRC Inspection Manual Chapter 0610*, "Power Reactor Inspection Reports," Appendix B, "Thresholds for Documentation."
The NRC determined that the described reportability determination weaknesses, if left uncorrected, could cause the same issues under the same conditions to become a more significant safety concern, due to the latent potential to fail to make a required report. The NRC determined that the issue did not apply to any specific cornerstone and was, therefore, not subject to the Significance Determination Process. The NRC also determined that the issue had the potential to impact the NRC's ability to perform its regulatory function, specifically, the ability of the NRC to monitor compliance with safety standards. Therefore, the NRC considered the issue to have extenuating circumstances that warranted documentation as a finding of No Color.
Inspection Report# : 2001008(pdf)
Significance: N/A Dec 18, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program 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. The licensee implemented corrective actions in a timely manner. Workers at the site expressed willingness to input safety issues into the problem identification and resolution program. However, the licensee's emergency diesel generator team, established to increase emphasis on emergency diesel generator reliability,
 
3Q/2002 Inspection Findings - River Bend 1                                                                    Page 18 of 18 was not involved in the recent operability assessment and subsequent root cause evaluations of an apparent concurrent inoperability of the Division I and II emergency diesel generators. This performance was not consistent with the expectation conveyed in a recent licensee response to agency concerns regarding the emergency diesel generator reliability at River Bend Station (Entergy Operations, Inc., Letter G9.5, G15.4.1 dated June 12, 2000). Further, the licensee's failure to make an early identification of the extent of condition was partly the result of a human performance error that reported the wrong piece of equipment as needing repair. The recent NRC Inspection Report 50-458/00-14 identified human performance errors as a cross-cutting finding at the site.
Inspection Report# : 2000018(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining human performance trend The inspectors identified a declining human performance trend with failure of personnel to adhere to plant procedural requirements or to maintain a questioning attitude as common elements. Approximately 27 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: N/A Nov 11, 2000 Identified By: NRC Item Type: FIN Finding Declining problem identification and resolution trend The inspectors identified a declining problem identification and resolution trend with not implementing timely corrective actions as a common element. Approximately 9 findings, which were documented as violations of NRC requirements during the previous 12 months, had a direct or credible impact on safety. This adverse performance trend is considered a cross-cutting finding not captured in individual findings.
Inspection Report# : 2000014(pdf)
Significance: SL-IV Nov 11, 2000 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized use of computer at an operations watch station Operations personnel inappropriately accessed nonjob related information on the operations shift superintendent's computer. The participation in potentially distracting activities at the operations shift superintendent's watch station was a violation of Technical Specification 5.4.1.a. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This finding was entered in the licensee's corrective action program as Condition Report 2000-1709. The inspectors determined that the safety significance of the potentially distracting activity at the operations shift superintendent's watch station was very low in that no actual plant problems occurred during the time in question which would have required the operations shift superintendent's response. The inspectors also determined that the finding was representative of an isolated human performance cross-cutting issue involving the failure to follow plant procedures.
Inspection Report# : 2000014(pdf)
Last modified : December 02, 2002
 
4Q/2002 Inspection Findings - River Bend 1                                                                                                Page 1 of 4 River Bend 1 Initiating Events Mitigating Systems Significance: TBD Sep 18, 2002 Identified By: Self Disclosing Item Type: AV Apparent Violation Failure to properly lock open condensate valve resulted in loss of feedwater flow following reactor scram.
(TBD) The inspectors identified an apparent violation of Technical Specification 5.4.1.a, which required that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33 lists the condensate system as one of the systems requiring operating procedures. System Operating Procedure SOP-0007, "Condensate System," Revision 21, required that Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 be locked open. On September 18, 2002, Valve VNM-FVC200 was found to be improperly locked in the open position. This failure to properly lock open CNM-FCV200 resulted in unexpected closure of the valve and a loss of feedwater flow to the reactor vessel following a reactor scram. The final significance of this issue will be determined using the Significance Determination Process.
Inspection Report# : 2002007(pdf)
Significance:        Aug 15, 2002 Identified By: NRC Item Type: FIN Finding Ineffective corrective actions caused station blackout diesel generator to be unavailable On August 15, 2002, the licensee performed a routine monthly performance test of the station blackout diesel generator. Four minutes into the one-hour run the diesel generator tripped on high coolant temperature. Similar failures of the station blackout diesel generator to run due to high temperature trips had occurred in each of the two previous monthly performance tests on June 21 and July 19, 2002. For each of these failures, the licensee identified an apparent cause for the failure and corrected the problems identified. Following the failure on August 15, 2002, the inspectors determined that the licensee-identified causes for the previous station blackout diesel generator failures were not accurate; therefore, the corrective actions taken were ineffective. The inspectors evaluated the ineffective corrective actions taken to correct two failures of the station blackout diesel generator using inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The inspectors determined that the finding was more than minor in that it affected the operability and availability of a risk-significant mitigating system, i.e., the station blackout diesel generator.
The inspectors determined that the failure to maintain the station blackout diesel generator operable was of very low safety significance (Green) because of the low likelihood of a station blackout event occurring, the probability that operators could restore the diesel following an initial failure, and the availability of all other standby electrical systems. This problem identification and resolution issue was entered into the licensee's corrective action program as CR-RBS-2002-0664.
Inspection Report# : 2002003(pdf)
Significance:        May 29, 2002 Identified By: NRC Item Type: FIN Finding Increased Division I Emergency Diesel Generator jacket cooling water leak rate caused diesel generator to be operable but degraded beyond the licensee's existing evaluation Following maintenance performed on May 9, 2002, to determine the source of a leak from the Division 1 emergency diesel generator jacket cooling water system, the leak rate more than doubled. The licensee's attempt to correct the problem on May 30, 2002, resulted in another increase in the leak rate to the point that makeup to the jacket cooling water system would be required within approximately 2 hours of Division I emergency diesel generator operation during a loss of offsite power. Although, the cause for the increased jacket water leak was repaired on June 4, 2002, the diesel generator remained degraded, but operable. The licensee planned to repair the original leak during the next extended diesel generator maintenance outage. The inspectors determined that the increased leak rate was beyond the licensee's evaluation that concluded that the Division 1 emergency diesel generator was degraded but operable. If left uncorrected, the jacket cooling water leak could have caused the emergency diesel generator to become inoperable and unavailable. The normal source of makeup water would not have been available during a loss of offsite power and the licensee did not develop a written procedure for use of an alternate makeup source until May 30, 2002. Using the significance determination process, the risk significance of the finding was determined to be very low because the emergency diesel generator remained operable, although degraded. This maintenance induced problem was documented in the licensee's corrective action program as Condition Report CR-RBS-2002-0672.
 
4Q/2002 Inspection Findings - River Bend 1                                                                                            Page 2 of 4 Inspection Report# : 2002002(pdf)
Significance:        May 12, 2002 Identified By: Self Disclosing Item Type: FIN Finding Operator action caused a high reactor water level trip of the running reactor feed pump following a planned scram from 26 percent power Following a planned reactor scram during a plant shutdown, operators failed to take manual control of the feedwater level control system in time to stop an unexpected rise in reactor water level until after the running reactor feed pump tripped on high reactor water level. The licensee determined that the reduction of the reactor pressure control setpoint and subsequent opening of the main turbine bypass valves caused a "swell" in reactor water level which contributed to the higher than expected reactor water level transient. The inspectors determined that the operators did not manually close and isolate one of the two automatic feedwater regulating valves in time to eliminate leakage past the feedwater regulating valve, and failed to reject water from the reactor through the reactor water cleanup system in time to stop the rise in reactor water level to the high level trip of the reactor feed pump. The failure of the operators to manually control reactor water level resulted in the unavailability of a risk-significant reactor feed pump. The inspectors, using the significance determination process, determined that the safety significance of the high reactor water level trip of the running reactor feed pump following a planned reactor scram was very low because the reactor feed pump was restarted from the main control room as soon as reactor water level was lowered and the high reactor water level trip signal was cleared, and other reactor water makeup sources remained available. This human performance error was documented in the licensee's corrective action program as Condition Report CR-RBS-2002-0688.
Inspection Report# : 2002002(pdf)
Significance:        May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding Station Blackout Diesel Generator inoperable due to discharged starting battery The station blackout diesel generator was found to be inoperable by the licensee because its starting battery had been allowed to completely discharge. The station blackout diesel generator had been moved from its normal storage location as a contingency for a planned maintenance outage on several Division I safety-related systems. The inspectors determined that the Division I maintenance outage contingency plan and the weekly work schedule did not plan for the return of the station blackout diesel generator to its normal storage location to re-energize its battery charger. The licensee determined that this is a repeat of a similar event of April 4, 1998, documented in Condition Report CR-RBS-1998-0384.
The failure to maintain its starting battery charged caused the risk significant station blackout diesel generator to be inoperable and unavailable.
The inspectors, using the significance determination process, determined that the safety significance of the unavailability of the station blackout diesel generator was very low because the length of time the diesel generator was unavailable was less than 24 hours and all other electrical systems were available during that time. This human performance error was documented is the licensee's corrective action program as Condition Report CR-RBS-2002-0664.
Inspection Report# : 2002002(pdf)
Barrier Integrity Emergency Preparedness Significance:        Mar 20, 2002 Identified By: NRC Item Type: VIO Violation Failure to develop a range of protective actions, disseminate emergency response information, and maintain the emergency plan for members of the public located within the owner controlled area The inspector identified one preliminary finding involving the failure to develop a range of protective actions, disseminate emergency response information, and maintain the emergency plan in accordance with the requirements of 10 CFR 50.54(q), planning standards §50.47(b)(10) and (7), and 10 CFR Part 50, Appendix E, section IV(G) pertaining to members of the public located in the owner controlled area. Three apparent violations are associated with the finding. The issues involved: (1) a failure to establish effective means or provisions for warning, advising, evacuating, and monitoring members of the public during an owner controlled area evacuation, (2) a failure to disseminate emergency response information to the public using facilities in the River Bend Station owner controlled area, and (3) a failure to update the emergency plan and procedures after the public was permitted access to facilities in the owner controlled area. The licensee has entered these issues into its
 
4Q/2002 Inspection Findings - River Bend 1                                                                                            Page 3 of 4 corrective action program in CR-RBS-2001-1713 and CR-RBS-2002-0183. This issue was preliminarily determined to have substantial safety significance (Yellow) because it represented a failure to meet a risk-significant emergency preparedness planning standard. UPDATE: On July 31, 2002, a Notice of Violation (EA-02-036) was issued regarding this issue. The violation was as follows: 10 CFR 50.54(q) strates, in part, that a licensee authorized to possess and operate a nuclear power reactor shall follow and maintain in effect emergency plans which meet the standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(7) requires that onsite emergency response plans for nuclear power reactors meet the following standard, which states, in part: "Information is made available to the public on a periodic basis on how they will be notified and what their initial actions should be in an emergency..." Contrary to the above, between 1985 and February 1, 2002, the licensee's emergency plan was not adequate to assure that information was made available to members of the public using River Bend Station's owner conrolled area regarding how members of the public would be notified of an evacuation order and what their initial actions should be in an emergency. Specifically, the licensee had not provided information to members of the public using the West Feliciana Community Development Foundation, the security firing range, the activity center, the outage campground, the Sportsman's Association base camp, and adjacent hunting and fishing areas in the owner controlled area about: (1) the process used to notify the public of an emergency, (2) circumstances under which the public in the licensee's owner controlled area would be directed to assembly and radiological monitoring stations, (3) the predetermined locations of the assembly and radiological monitoring stations, (4) evacuation routes to the predetermined assembly and radiological monitoring stations, and (5) the radiological monitoring and decontamination process. This violation is associated with a White Significance Determination Process finding. The NRC performed this supplemental inspection to assess the licensee's evaluation associated with the failure to meet the requirements of 10 CFR 50.54(q), in that the licensee did not follow and maintain emergency plans and procedures which meet the standards in 10 CFR 50.47(b)(7). This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report 50-458/2002-05. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector noted that although some weaknesses in the root cause analysis were apparent, the licensee performed a comprehensive evaluation of the White finding. The licensee's evaluation identified the primary root causes of the performance issue to be inadequate implementation of the public information program and inadequate 10 CFR 50.54(q) evaluations. 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 2002, therefore it will no longer be considered in assessing plant performance after the fourth quarter of 2002. Supplemental Inspection documented in NRC IR 50-458/02-08.
Inspection Report# : 2002005(pdf)
Significance:        Mar 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Reduction in Emergency Plan Effectiveness without Prior NRC Approval This noncited violation is described in a letter to the licensee dated July 31, 2002, regarding the "Final Significance Determination for a White Finding and Notice of Violation." Green. A noncited violation of very low risk significance was identified for failure to comply with the requirements of 10 CFR 50.54(q). Between 1985 and January 2002, the licensee reduced the effectiveness of its emergency plan without Commission approval when it: (1) changed from the use of security vehicles equipped with permanently-mounted public address systems to the use of vehicles without such systems, and relied on portable public address systems stored onsite, (2) canceled emergency plan implementing procedure EIP-2-026, "Evacuation, Personnel Accountability, and Search and Rescue," Revision 11, and (3) permitted several changes in the public's use of the River Bend Station owner controlled area without evaluation of the impact of those changes on the emergency plan. 10 CFR 50.54(q) requires, in part, that each nuclear power plant licensee may make changes to its emergency plans without Commission approval only if the changes do not decrease the effectiveness of the plans and the plans, as changed, continue to meet the standards of 10 CFR 50.47(b) and the requirements of Appendix E of 10 CFR Part 50. The decrease in effectiveness of the emergency plan resulting from the failure to evaluate changes in the station owner controlled area, changes to emergency plan implementing procedures, and changes in emergency notification methods used by security officers, was a performance deficiency. The finding was more than minor because it was associated with one of the Emergency Preparedness cornerstone attributes (Plan Changes) and affected the associated cornerstone objective. Using the Emergency Preparedness Significance Determination Process, the inspector determined the violation had very low risk significance because the violation did not constitute a failure to meet an emergency planning standard as defined by 10 CFR 50.47(b). Because of the very low safety significance and because the licensee included the finding in their corrective action program as Condition Report 2002-0183, this finding is being treated as a noncited violation in accordance with Section VI.A of the NRC Enforcement Policy.
Inspection Report# : 2002005(pdf)
Occupational Radiation Safety Public Radiation Safety
 
4Q/2002 Inspection Findings - River Bend 1 Page 4 of 4 Physical Protection Miscellaneous Last modified : March 25, 2003
 
1Q/2003 Inspection Findings - River Bend 1                                                                      Page 1 of 4 River Bend 1 1Q/2003 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to take proper corrective actions for low pressure core spray pump minimum flow valve failure resulted in the failure of the residual heat removal pump minimum flow valve The inspectors identified a noncited violation of 10 CFR 50 Appendix B Criterion XVI for failure to take proper corrective action following a failure of the low pressure core spray pump minimum flow valve that resulted in an identical failure of the residual heat removal Pump A minimum flow valve nine months later. The inspector identified non-cited violation was greater than minor because it was associated with the mitigating systems cornerstone objective to ensure the availability, reliability and capability of systems (residual heat removal Train A) that respond to initiating events to prevent undesirable consequences. With the minimum flow valve open, residual heat removal Train A was not able to meet its design flow rate for either the low pressure coolant injection or suppression pool cooling mode of system operation. The inspectors evaluated the finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Reactors" and determined that the residual heat removal Pump A minimum flow valve failure was of very low safety significance because the other low pressure coolant injection systems were available and the other train of suppression pool cooling was available at the time.
Inspection Report# : 2003003(pdf)
Significance: TBD Sep 18, 2002 Identified By: Self Disclosing Item Type: AV Apparent Violation Failure to properly lock open condensate valve resulted in loss of feedwater flow following reactor scram.
(TBD) The inspectors identified an apparent violation of Technical Specification 5.4.1.a, which required that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33 lists the condensate system as one of the systems requiring operating procedures. System Operating Procedure SOP-0007, "Condensate System," Revision 21, required that Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 be locked open. On September 18, 2002, Valve VNM-FVC200 was found to be improperly locked in the open position. This failure to properly lock open CNM-FCV200 resulted in unexpected closure of the valve and a loss of feedwater flow to the reactor vessel following a reactor scram. The final significance of this issue will be determined using the Significance Determination Process.
Inspection Report# : 2002007(pdf)
Significance:        Aug 15, 2002 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - River Bend 1                                                                        Page 2 of 4 Item Type: FIN Finding Ineffective corrective actions caused station blackout diesel generator to be unavailable On August 15, 2002, the licensee performed a routine monthly performance test of the station blackout diesel generator. Four minutes into the one-hour run the diesel generator tripped on high coolant temperature. Similar failures of the station blackout diesel generator to run due to high temperature trips had occurred in each of the two previous monthly performance tests on June 21 and July 19, 2002. For each of these failures, the licensee identified an apparent cause for the failure and corrected the problems identified. Following the failure on August 15, 2002, the inspectors determined that the licensee-identified causes for the previous station blackout diesel generator failures were not accurate; therefore, the corrective actions taken were ineffective. The inspectors evaluated the ineffective corrective actions taken to correct two failures of the station blackout diesel generator using inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The inspectors determined that the finding was more than minor in that it affected the operability and availability of a risk-significant mitigating system, i.e., the station blackout diesel generator. The inspectors determined that the failure to maintain the station blackout diesel generator operable was of very low safety significance (Green) because of the low likelihood of a station blackout event occurring, the probability that operators could restore the diesel following an initial failure, and the availability of all other standby electrical systems. This problem identification and resolution issue was entered into the licensee's corrective action program as CR-RBS-2002-0664.
Inspection Report# : 2002003(pdf)
Significance:      May 29, 2002 Identified By: NRC Item Type: FIN Finding Increased Division I Emergency Diesel Generator jacket cooling water leak rate caused diesel generator to be operable but degraded beyond the licensee's existing evaluation Following maintenance performed on May 9, 2002, to determine the source of a leak from the Division 1 emergency diesel generator jacket cooling water system, the leak rate more than doubled. The licensee's attempt to correct the problem on May 30, 2002, resulted in another increase in the leak rate to the point that makeup to the jacket cooling water system would be required within approximately 2 hours of Division I emergency diesel generator operation during a loss of offsite power. Although, the cause for the increased jacket water leak was repaired on June 4, 2002, the diesel generator remained degraded, but operable. The licensee planned to repair the original leak during the next extended diesel generator maintenance outage. The inspectors determined that the increased leak rate was beyond the licensee's evaluation that concluded that the Division 1 emergency diesel generator was degraded but operable. If left uncorrected, the jacket cooling water leak could have caused the emergency diesel generator to become inoperable and unavailable. The normal source of makeup water would not have been available during a loss of offsite power and the licensee did not develop a written procedure for use of an alternate makeup source until May 30, 2002. Using the significance determination process, the risk significance of the finding was determined to be very low because the emergency diesel generator remained operable, although degraded. This maintenance induced problem was documented in the licensee's corrective action program as Condition Report CR-RBS-2002-0672.
Inspection Report# : 2002002(pdf)
Significance:      May 12, 2002 Identified By: Self Disclosing Item Type: FIN Finding Operator action caused a high reactor water level trip of the running reactor feed pump following a planned scram from 26 percent power Following a planned reactor scram during a plant shutdown, operators failed to take manual control of the feedwater level control system in time to stop an unexpected rise in reactor water level until after the running reactor feed pump file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            07/22/2003
 
1Q/2003 Inspection Findings - River Bend 1                                                                        Page 3 of 4 tripped on high reactor water level. The licensee determined that the reduction of the reactor pressure control setpoint and subsequent opening of the main turbine bypass valves caused a "swell" in reactor water level which contributed to the higher than expected reactor water level transient. The inspectors determined that the operators did not manually close and isolate one of the two automatic feedwater regulating valves in time to eliminate leakage past the feedwater regulating valve, and failed to reject water from the reactor through the reactor water cleanup system in time to stop the rise in reactor water level to the high level trip of the reactor feed pump. The failure of the operators to manually control reactor water level resulted in the unavailability of a risk-significant reactor feed pump. The inspectors, using the significance determination process, determined that the safety significance of the high reactor water level trip of the running reactor feed pump following a planned reactor scram was very low because the reactor feed pump was restarted from the main control room as soon as reactor water level was lowered and the high reactor water level trip signal was cleared, and other reactor water makeup sources remained available. This human performance error was documented in the licensee's corrective action program as Condition Report CR-RBS-2002-0688.
Inspection Report# : 2002002(pdf)
Significance:      May 11, 2002 Identified By: Self Disclosing Item Type: FIN Finding Station Blackout Diesel Generator inoperable due to discharged starting battery The station blackout diesel generator was found to be inoperable by the licensee because its starting battery had been allowed to completely discharge. The station blackout diesel generator had been moved from its normal storage location as a contingency for a planned maintenance outage on several Division I safety-related systems. The inspectors determined that the Division I maintenance outage contingency plan and the weekly work schedule did not plan for the return of the station blackout diesel generator to its normal storage location to re-energize its battery charger. The licensee determined that this is a repeat of a similar event of April 4, 1998, documented in Condition Report CR-RBS-1998-0384. The failure to maintain its starting battery charged caused the risk significant station blackout diesel generator to be inoperable and unavailable. The inspectors, using the significance determination process, determined that the safety significance of the unavailability of the station blackout diesel generator was very low because the length of time the diesel generator was unavailable was less than 24 hours and all other electrical systems were available during that time. This human performance error was documented is the licensee's corrective action program as Condition Report CR-RBS-2002-0664.
Inspection Report# : 2002002(pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - River Bend 1            Page 4 of 4 Physical Protection Miscellaneous Last modified : May 30, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html 07/22/2003
 
2Q/2003 Inspection Findings - River Bend 1                                                                        Page 1 of 6 River Bend 1 2Q/2003 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Jun 10, 2003 Identified By: Self Disclosing Item Type: FIN Finding Foreign material caused failure of one residual heat removal equipment room floor drain sump pump while the other pump was unavailable The inspectors identified a self-revealing finding for failure to control foreign material in the residual heat removal Train B equipment room which resulted in the failure of one of two floor drain pumps while the other floor drain pump was unavailable. The finding was of very low safety significance because the floor drain sump pump failure did not cause an actual loss of safety function for residual heat removal Train B. The inspectors determined that the licensee's failure to control foreign material in the residual heat removal Train B equipment room, which resulted in the fouling and unavailability of floor drain Pump DFR-P3L while Pump DFR-P3E was also unavailable, was a performance deficiency. This self-revealing finding was more than minor because, if left uncorrected and a leak developed in the residual heat removal Train B equipment room, the unavailability of both floor drain sump pumps could lead to a loss of residual heat removal Train B. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the phase one screening of the finding, the inspectors determined that the finding was of very low safety significance because the floor drain sump pump failure did not increase the likelihood of a plant trip or degrade more than one train of any safety system. The finding is documented in the licensee's corrective action program as CR-RBS-2003-2368.
Inspection Report# : 2003004(pdf)
Significance:        Feb 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain watertight integrity of severe weather doors compromised the availability of standby service water system The inspectors identified a noncited violation for the failure of the licensee to comply with 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation was for failure to incorporate necessary measures into station procedures to ensure that the design basis of the doors at the end of underground G-Tunnel was maintained. The finding was more than minor because it was associated with flood protection measures and degraded the ability to meet the mitigating systems cornerstone objective. It had an adverse impact on the flooding potential of the G-Tunnel, which opened into the base of the standby cooling tower, and challenged the availability of the standby service water system.
The finding is of very low safety significance because of the existing condition of the door seals, the availability of two nonsafety-related sump pumps at the base of the standby cooling tower, the relative height of the control circuits and motor operators of the cooling tower inlet valves and the possibility of operator action to manually initiate standby file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - River Bend 1                                                                      Page 2 of 6 service water before the failure of the standby cooling tower inlet valves. The finding was documented in the licensee's corrective action program as CR-RBS-2003-1894.
Inspection Report# : 2003004(pdf)
Significance:        Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to take proper corrective actions for low pressure core spray pump minimum flow valve failure resulted in the failure of the residual heat removal pump minimum flow valve The inspectors identified a noncited violation of 10 CFR 50 Appendix B Criterion XVI for failure to take proper corrective action following a failure of the low pressure core spray pump minimum flow valve that resulted in an identical failure of the residual heat removal Pump A minimum flow valve nine months later. The inspector identified non-cited violation was greater than minor because it was associated with the mitigating systems cornerstone objective to ensure the availability, reliability and capability of systems (residual heat removal Train A) that respond to initiating events to prevent undesirable consequences. With the minimum flow valve open, residual heat removal Train A was not able to meet its design flow rate for either the low pressure coolant injection or suppression pool cooling mode of system operation. The inspectors evaluated the finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Reactors" and determined that the residual heat removal Pump A minimum flow valve failure was of very low safety significance because the other low pressure coolant injection systems were available and the other train of suppression pool cooling was available at the time.
Inspection Report# : 2003003(pdf)
Significance: TBD Sep 18, 2002 Identified By: Self Disclosing Item Type: AV Apparent Violation Failure to properly lock open condensate valve resulted in loss of feedwater flow following reactor scram.
(TBD) The inspectors identified an apparent violation of Technical Specification 5.4.1.a, which required that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33 lists the condensate system as one of the systems requiring operating procedures. System Operating Procedure SOP-0007, "Condensate System," Revision 21, required that Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 be locked open. On September 18, 2002, Valve VNM-FVC200 was found to be improperly locked in the open position. This failure to properly lock open CNM-FCV200 resulted in unexpected closure of the valve and a loss of feedwater flow to the reactor vessel following a reactor scram. The final significance of this issue will be determined using the Significance Determination Process.
Inspection Report# : 2002007(pdf)
Significance:        Aug 15, 2002 Identified By: NRC Item Type: FIN Finding Ineffective corrective actions caused station blackout diesel generator to be unavailable On August 15, 2002, the licensee performed a routine monthly performance test of the station blackout diesel generator. Four minutes into the one-hour run the diesel generator tripped on high coolant temperature. Similar failures of the station blackout diesel generator to run due to high temperature trips had occurred in each of the two previous monthly performance tests on June 21 and July 19, 2002. For each of these failures, the licensee identified an apparent cause for the failure and corrected the problems identified. Following the failure on August 15, 2002, the inspectors determined that the licensee-identified causes for the previous station blackout diesel generator failures were not accurate; therefore, the corrective actions taken were ineffective. The inspectors evaluated the ineffective corrective file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - River Bend 1                                                                        Page 3 of 6 actions taken to correct two failures of the station blackout diesel generator using inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The inspectors determined that the finding was more than minor in that it affected the operability and availability of a risk-significant mitigating system, i.e., the station blackout diesel generator. The inspectors determined that the failure to maintain the station blackout diesel generator operable was of very low safety significance (Green) because of the low likelihood of a station blackout event occurring, the probability that operators could restore the diesel following an initial failure, and the availability of all other standby electrical systems. This problem identification and resolution issue was entered into the licensee's corrective action program as CR-RBS-2002-0664.
Inspection Report# : 2002003(pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance:      Apr 09, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses ALARA that were associated with RWP 2003-1929 A finding was identified because the licensee failed to maintain collective doses ALARA. Specifically, the work activity collective dose associated with RWP 2003-1929 Task 1, "Refueling Outage 11 Recirculation Pump Work,"
exceeded 5 person-rem and exceeded the dose estimation by more than 50 percent. The failure to maintain collective doses ALARA is a performance deficiency. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/projected dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). This occurrence involved worker inefficiencies, inadequate planning, scheduling and supervisory oversight which resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this finding was found to have no more than very low safety significance because the finding was an ALARA Planning issue, the licensee's three-year rolling average collective dose was greater than 240 person-rem, the actual dose for the work activity was not more than 25 person-rem, and there were no more than four occurrences.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses ALARA that were associated with RWP 2003-1935 A finding was identified because the licensee failed to maintain collective doses ALARA. Specifically, the work file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            10/08/2003
 
2Q/2003 Inspection Findings - River Bend 1                                                                    Page 4 of 6 activity collective dose associated with RWP 2003-1935, "Drywell Valve Maintenance, to include Repacks and Support Work," exceeded 5 person-rem and exceeded the dose estimation by more than 50 percent. The failure to maintain collective doses ALARA is a performance deficiency. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/projected dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). This occurrence involved worker inefficiencies and inadequate planning which resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this finding was found to have no more than very low safety significance because the finding was an ALARA Planning issue, the licensee's three-year rolling average collective dose was greater than 240 person-rem, the actual dose for the work activity was not more than 25 person-rem, and there were no more than four occurrences.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to develop a sufficiently detailed work plan The licensee failed to develop a sufficiently detailed work plan for the decontamination of the reactor vessel bellows, in violation of Technical Specification 5.4.1. a. The work plan failed to provide guidance on maintaining highly contaminated surfaces (the reactor vessel bellows surface) wet, using a hydrolaser with a rotary surface cleaner, or briefing the individual using the hydrolaser. The lack of a detailed work planned contributed to an unexpected increase in airborne radioactivity and unplanned personnel exposures. This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an as low as is reasonably achievable (ALARA) finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to survey The licensee failed, in three examples, to survey or evaluate radiological hazards when conducting reactor vessel bellows decontamination, in violation of 10 CFR 20.1501(a). First, the licensee failed to evaluate the highest concentration of radioactive contamination on the reactor vessel bellows. Additionally, the licensee failed to evaluate the airborne radioactivity in the immediate vicinity of reactor vessel bellows contamination. Later, the licensee failed to evaluate the airborne radioactivity levels throughout the containment building when continuous air monitors alarmed.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                        10/08/2003
 
2Q/2003 Inspection Findings - River Bend 1                                                                      Page 5 of 6 reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to post an airborne radioactivity area The licensee failed to post the reactor containment building as an airborne radioactivity area, in violation with 10 CFR 20.1902(d). Airborne radioactivity levels exceeded the allowable limits in 10 CFR Part 20, Appendix B by as much as 3.5 times. The condition existed for at least five hours. This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to instruct workers Following an occurrence that caused an airborne radioactivity area, the licensee failed to inform workers of radiological conditions that had changed and of precautions to minimize exposure, in violation of 10 CFR 19.12. This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to control a locked high radiation area The licensee failed to control an area with dose rates of 1000 millirems per hour as a locked high radiation area, in violation of Technical Specification 5.7.2. After a plant scram on September 18, 2002, a worker entered the reactor core isolation cooling area on the 95-foot elevation of the auxiliary building and received an electronic dosimeter dose file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - River Bend 1                                                                      Page 6 of 6 rate alarm. A crud burst resulting from a transient that occurred approximately three hours previously caused the dose levels in the area entered by the worker to increase to 1000 millirems per hour. Historically, the site has experienced crud bursts under similar conditions and the increase in dose rate should have been anticipated and evaluated. This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Public Radiation Safety Physical Protection Miscellaneous Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          10/08/2003
 
3Q/2003 Inspection Findings - River Bend 1                                                                      Page 1 of 8 River Bend 1 3Q/2003 Plant Inspection Findings Initiating Events Significance: TBD Sep 01, 2003 Identified By: Self Disclosing Item Type: AV Apparent Violation Human performance error results in air bound normal service water pump The inspectors identified an apparent self-revealing violation of Technical Specification 5.4.1.a, the significance of which has yet to be determined. A human performance error caused the isolation of the air release valve for normal service water Pump C. The air release valve for a normal service water pump served as a high point vent on the system while the pump was secured. As a result, normal service water Pump C became air bound while in standby, and failed to develop discharge pressure when started during a manual swap of running normal service water pumps on September 1, 2003.
The inspectors determined that the failure to maintain normal service water Pump C discharge air release valve isolation Valve SWP-V3312C open was an apparent violation of normal service water system operating procedure SOP-0018, Attachment 1A, "Valve Lineup - Normal Service Water," Revision 32. The human performance error was more than minor because it was associated with an increase in the likelihood of an initiating event. The inspectors reviewed this finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The result of the phase one screening process and the inspectors' review of the increased likelihood of a loss of normal service water was referral of the issue to the regional senior reactor analyst for further review and risk determination.
Inspection Report# : 2003005(pdf)
Mitigating Systems Significance:      Jun 17, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation HPCS Inadvertently Disabled Due to Personnel Error During Installation of Clearance Order The inspectors identified a self-revealing violation for failure to comply with Technical Specification 5.4.1.a. Operators mistakenly racked out the HPCS pump breaker when implementing a clearance order on a standby service water.
This self-revealing finding was more than minor because the HPCS safety function was made unavailable. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The inspectors determined that the finding was of very low safety significance (Green) because the HPCS pump was not functional for less than one hour. Recovery credit was given for operator actions necessary to restore the equipment lineup and recover the safety function.
Inspection Report# : 2003005(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - River Bend 1                                                                        Page 2 of 8 Significance:      Jun 10, 2003 Identified By: Self Disclosing Item Type: FIN Finding Foreign material caused failure of one residual heat removal equipment room floor drain sump pump while the other pump was unavailable The inspectors identified a self-revealing finding for failure to control foreign material in the residual heat removal Train B equipment room which resulted in the failure of one of two floor drain pumps while the other floor drain pump was unavailable. The finding was of very low safety significance because the floor drain sump pump failure did not cause an actual loss of safety function for residual heat removal Train B.
The inspectors determined that the licensee's failure to control foreign material in the residual heat removal Train B equipment room, which resulted in the fouling and unavailability of floor drain Pump DFR-P3L while Pump DFR-P3E was also unavailable, was a performance deficiency. This self-revealing finding was more than minor because, if left uncorrected and a leak developed in the residual heat removal Train B equipment room, the unavailability of both floor drain sump pumps could lead to a loss of residual heat removal Train B. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the phase one screening of the finding, the inspectors determined that the finding was of very low safety significance because the floor drain sump pump failure did not increase the likelihood of a plant trip or degrade more than one train of any safety system. The finding is documented in the licensee's corrective action program as CR-RBS-2003-2368.
Inspection Report# : 2003004(pdf)
Significance:      May 07, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Automatic Initiation of Standby Service Water System Due to Inadequate Control of System Operation The inspectors identified a self-revealing violation of Technical Specification 5.4.1 because operators lined up service water to the reactor plant and turbine plant cooling water systems such that an automatic start of standby service water occurred on low system pressure while shifting normal service water pumps. Three heat exchangers in each system were in service when the system operating procedures for reactor plant and turbine plant cooling water allow only two heat exchangers in operation per system.
This finding is greater than minor because it was associated with the ability to meet the mitigating systems cornerstone objective and because a plant transient occurred. The inspectors determined that the finding was of very low safety significance (Green), since the finding did not represent an actual loss of safety function of a single train.
Inspection Report# : 2003005(pdf)
Significance:      Feb 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain watertight integrity of severe weather doors compromised the availability of standby service water system The inspectors identified a noncited violation for the failure of the licensee to comply with 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation was for failure to incorporate necessary measures into station procedures to ensure that the design basis of the doors at the end of underground G-Tunnel was maintained.
The finding was more than minor because it was associated with flood protection measures and degraded the ability to file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - River Bend 1                                                                      Page 3 of 8 meet the mitigating systems cornerstone objective. It had an adverse impact on the flooding potential of the G-Tunnel, which opened into the base of the standby cooling tower, and challenged the availability of the standby service water system. The finding is of very low safety significance because of the existing condition of the door seals, the availability of two nonsafety-related sump pumps at the base of the standby cooling tower, the relative height of the control circuits and motor operators of the cooling tower inlet valves and the possibility of operator action to manually initiate standby service water before the failure of the standby cooling tower inlet valves. The finding was documented in the licensee's corrective action program as CR-RBS-2003-1894.
Inspection Report# : 2003004(pdf)
Significance:        Dec 28, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to take proper corrective actions for low pressure core spray pump minimum flow valve failure resulted in the failure of the residual heat removal pump minimum flow valve The inspectors identified a noncited violation of 10 CFR 50 Appendix B Criterion XVI for failure to take proper corrective action following a failure of the low pressure core spray pump minimum flow valve that resulted in an identical failure of the residual heat removal Pump A minimum flow valve nine months later.
The inspector identified non-cited violation was greater than minor because it was associated with the mitigating systems cornerstone objective to ensure the availability, reliability and capability of systems (residual heat removal Train A) that respond to initiating events to prevent undesirable consequences. With the minimum flow valve open, residual heat removal Train A was not able to meet its design flow rate for either the low pressure coolant injection or suppression pool cooling mode of system operation. The inspectors evaluated the finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Reactors" and determined that the residual heat removal Pump A minimum flow valve failure was of very low safety significance because the other low pressure coolant injection systems were available and the other train of suppression pool cooling was available at the time.
Inspection Report# : 2003003(pdf)
Significance: TBD Sep 18, 2002 Identified By: Self Disclosing Item Type: AV Apparent Violation Failure to properly lock open condensate valve resulted in loss of feedwater flow following reactor scram.
(TBD) The inspectors identified an apparent violation of Technical Specification 5.4.1.a, which required that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33 lists the condensate system as one of the systems requiring operating procedures.
System Operating Procedure SOP-0007, "Condensate System," Revision 21, required that Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 be locked open. On September 18, 2002, Valve VNM-FVC200 was found to be improperly locked in the open position. This failure to properly lock open CNM-FCV200 resulted in unexpected closure of the valve and a loss of feedwater flow to the reactor vessel following a reactor scram.
The final significance of this issue will be determined using the Significance Determination Process.
Inspection Report# : 2002007(pdf)
Barrier Integrity file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - River Bend 1                                                                    Page 4 of 8 Emergency Preparedness Occupational Radiation Safety Significance:      Aug 22, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses associated with RWP 2003-1800 ALARA The inspector identified an ALARA finding because performance deficiencies resulted in a collective dose of the work activity that exceeded 5 person-rem and exceeded the legitimate dose estimation by more than 50 percent. Specifically, Radiation Work Permit 2003-1800, "RF-11 Refueling Activities," accrued 34.962 person-rem and exceeded the dose estimate (19.939 person-rem) by 75 percent. A primary cause for the unplanned dose was the licensee's failure to effectively schedule the use of the Alternate Heat Decay Removal System, a system which had previously proven to be effective at removing radioactivity from the refueling pool. The licensee also failed to limit the number of personnel on the refueling bridge to the planned number, thus causing the work activity to accrue more collective dose than estimated. A contamination incident during the disassembly of the reactor vessel was caused by poor planning and required additional time for cleanup.
This finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/estimated dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). The finding involved a failure to maintain or implement, to the extent practical, procedures or engineering controls needed to achieve occupational doses that were ALARA, and that resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this ALARA finding was found to have no more than very low safety significance because the licensee's 3-year rolling average collective dose was not greater than 240 person-rem.
Inspection Report# : 2003005(pdf)
Significance:      Aug 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a radiological hazard The team identified a non-cited violation of Technical Specification 5.4.1a because the licensee failed to post a radiological hazard (hot spot). Station Procedure RP-109, "Hot Spot Program," Revision 0, Step 5.2.1, required that hot spots are identified with a hot spot tag to alert workers of the hazard. However, on August 19, 2003, the team identified a hot spot on an accessible drain line from the radwaste sample sink reading 200 millirem per hour on contact and 50 millirem per hour at one foot from the source. The licensee performed a survey 11 days earlier that identified the radiation levels, however, the technician and the survey reviewer failed to tag the hot spot to warn workers of the hazard.
The finding was more than minor because it was associated with the Occupational Radiation Safety cornerstone attribute (Program and Process) and affected the associated cornerstone objective. The finding involved the potential for a workers unplanned or unintended dose resulting from actions contrary to procedures. When processed through the file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                        01/12/2004
 
3Q/2003 Inspection Findings - River Bend 1                                                                    Page 5 of 8 Occupational Radiation Safety Significance Determination Process the team determined that the finding had very low safety significance because the finding did not involve as low as is reasonably achievable (ALARA) planning or work controls, no individual received an overexposure or a substantial potential for overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2003009(pdf)
Significance:      Apr 09, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses ALARA that were associated with RWP 2003-1929 A finding was identified because the licensee failed to maintain collective doses ALARA. Specifically, the work activity collective dose associated with RWP 2003-1929 Task 1, "Refueling Outage 11 Recirculation Pump Work,"
exceeded 5 person-rem and exceeded the dose estimation by more than 50 percent.
The failure to maintain collective doses ALARA is a performance deficiency. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/projected dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). This occurrence involved worker inefficiencies, inadequate planning, scheduling and supervisory oversight which resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this finding was found to have no more than very low safety significance because the finding was an ALARA Planning issue, the licensee's three-year rolling average collective dose was greater than 240 person-rem, the actual dose for the work activity was not more than 25 person-rem, and there were no more than four occurrences.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses ALARA that were associated with RWP 2003-1935 A finding was identified because the licensee failed to maintain collective doses ALARA. Specifically, the work activity collective dose associated with RWP 2003-1935, "Drywell Valve Maintenance, to include Repacks and Support Work," exceeded 5 person-rem and exceeded the dose estimation by more than 50 percent.
The failure to maintain collective doses ALARA is a performance deficiency. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/projected dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). This occurrence involved worker inefficiencies and inadequate planning which resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this finding was found to have no more than very low safety significance because the finding was an ALARA Planning issue, the licensee's three-year rolling average collective dose was greater than 240 person-rem, the actual dose for the work activity was not more than 25 person-rem, and there were no more than four occurrences.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                        01/12/2004
 
3Q/2003 Inspection Findings - River Bend 1                                                                      Page 6 of 8 Failure to develop a sufficiently detailed work plan The licensee failed to develop a sufficiently detailed work plan for the decontamination of the reactor vessel bellows, in violation of Technical Specification 5.4.1. a. The work plan failed to provide guidance on maintaining highly contaminated surfaces (the reactor vessel bellows surface) wet, using a hydrolaser with a rotary surface cleaner, or briefing the individual using the hydrolaser. The lack of a detailed work planned contributed to an unexpected increase in airborne radioactivity and unplanned personnel exposures.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an as low as is reasonably achievable (ALARA) finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to survey The licensee failed, in three examples, to survey or evaluate radiological hazards when conducting reactor vessel bellows decontamination, in violation of 10 CFR 20.1501(a). First, the licensee failed to evaluate the highest concentration of radioactive contamination on the reactor vessel bellows. Additionally, the licensee failed to evaluate the airborne radioactivity in the immediate vicinity of reactor vessel bellows contamination. Later, the licensee failed to evaluate the airborne radioactivity levels throughout the containment building when continuous air monitors alarmed.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to post an airborne radioactivity area The licensee failed to post the reactor containment building as an airborne radioactivity area, in violation with 10 CFR 20.1902(d). Airborne radioactivity levels exceeded the allowable limits in 10 CFR Part 20, Appendix B by as much as 3.5 times. The condition existed for at least five hours.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - River Bend 1                                                                      Page 7 of 8 This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to instruct workers Following an occurrence that caused an airborne radioactivity area, the licensee failed to inform workers of radiological conditions that had changed and of precautions to minimize exposure, in violation of 10 CFR 19.12.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to control a locked high radiation area The licensee failed to control an area with dose rates of 1000 millirems per hour as a locked high radiation area, in violation of Technical Specification 5.7.2. After a plant scram on September 18, 2002, a worker entered the reactor core isolation cooling area on the 95-foot elevation of the auxiliary building and received an electronic dosimeter dose rate alarm. A crud burst resulting from a transient that occurred approximately three hours previously caused the dose levels in the area entered by the worker to increase to 1000 millirems per hour. Historically, the site has experienced crud bursts under similar conditions and the increase in dose rate should have been anticipated and evaluated.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - River Bend 1                                                                      Page 8 of 8 compromise the licensee's ability to assess dose, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Public Radiation Safety Significance:        Aug 22, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to properly control radioactive material A self-revealing non-cited violation of Technical Specification 5.4.1a was reviewed by the team because the licensee did not prevent the release of detectable licensed radioactive material from the controlled access area. Specifically, Section 5.1.1 of Procedure RSP-213, "Control and Handling of Radioactive Materials," Revision 16, stated, in part, that material can be unconditionally released from the controlled access area if there is no detectable loose surface and fixed contamination above background radiation levels. However, on March 31, 2003, the licensee failed to evaluate an item, against their procedural criteria, prior to it being unconditionally released from the controlled access area and subsequently released from the protected area. Fixed contamination levels were as high as 1,000 corrected counts per minute per probe area.
The finding was more than minor because it was associated with the Public Radiation Safety cornerstone attribute (Program and Process) and affected the associated cornerstone objective. The finding involved an occurrence in the radioactive material control program that was contrary to licensee procedures. When processed through the Public Radiation Safety Significance Determination Process, the team determined the finding had very low safety significance because the public exposure associated with the item was less than 5 millirem and there were not more than 5 occurrences.
Inspection Report# : 2003009(pdf)
Physical Protection Significance: N/A Aug 07, 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 Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          01/12/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                      Page 1 of 9 River Bend 1 4Q/2003 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding Normal Service Water pump found to be air-bound when called upon to run A self-revealing finding was identified and determined to be of very low safety significance. A human performance error caused the isolation of the air release valve for normal service water Pump C. The air release valve for a normal service water pump served as a high point vent on the system while the pump was secured. As a result, normal service water Pump C became air bound while in standby, and failed to develop discharge pressure when started during a manual swap of running normal service water pumps on September 1, 2003. The inspectors determined that the finding did not represent a noncompliance because it occurred on non-safety-related normal service water system.
The inspectors determined that the failure to maintain normal service water Pump C discharge air release valve isolation valve open was more than minor because it was associated with an increase in the likelihood of an initiating event. The finding was of very low safety significance because there was only a small increase in the likelihood of a loss of normal service water with one of the three 50 percent capacity normal service water pumps unavailable and because the standby service water system was available throughout the time normal service Pump C was air bound.
Inspection Report# : 2003006(pdf)
Significance:      Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding Repeat failure of main turbine control system hydraulic lines leads to manual reactor scram and turbine trip A self-revealing finding was identified for failure to properly diagnose a failure of turbine control hydraulic line failure in August 2000. As a result, a similar line failed on February 22, 2003, causing the operators to scram the reactor and trip the main turbine. The inspectors determined that the finding did not represent a noncompliance because it occurred on non-safety-related secondary plant equipment.
The inspectors determined that this problem identification and resolution finding is more than minor because the misdiagnosis of the August 31, 2000 failure contributed to the February 22, 2003, failure. The finding affected the initiating events cornerstone and was considered to have very low safety significance because it did not contribute to the likelihood of a LOCA, nor the likelihood of both a reactor scram and mitigating equipment or functions being unavailable, and because there was no increased likelihood of a fire or internal/external flood.
Inspection Report# : 2003006(pdf)
Mitigating Systems file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                        Page 2 of 9 Significance:      Jun 17, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation HPCS Inadvertently Disabled Due to Personnel Error During Installation of Clearance Order The inspectors identified a self-revealing violation for failure to comply with Technical Specification 5.4.1.a. Operators mistakenly racked out the HPCS pump breaker when implementing a clearance order on a standby service water.
This self-revealing finding was more than minor because the HPCS safety function was made unavailable. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The inspectors determined that the finding was of very low safety significance (Green) because the HPCS pump was not functional for less than one hour. Recovery credit was given for operator actions necessary to restore the equipment lineup and recover the safety function.
Inspection Report# : 2003005(pdf)
Significance:      Jun 10, 2003 Identified By: Self Disclosing Item Type: FIN Finding Foreign material caused failure of one residual heat removal equipment room floor drain sump pump while the other pump was unavailable The inspectors identified a self-revealing finding for failure to control foreign material in the residual heat removal Train B equipment room which resulted in the failure of one of two floor drain pumps while the other floor drain pump was unavailable. The finding was of very low safety significance because the floor drain sump pump failure did not cause an actual loss of safety function for residual heat removal Train B.
The inspectors determined that the licensee's failure to control foreign material in the residual heat removal Train B equipment room, which resulted in the fouling and unavailability of floor drain Pump DFR-P3L while Pump DFR-P3E was also unavailable, was a performance deficiency. This self-revealing finding was more than minor because, if left uncorrected and a leak developed in the residual heat removal Train B equipment room, the unavailability of both floor drain sump pumps could lead to a loss of residual heat removal Train B. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the phase one screening of the finding, the inspectors determined that the finding was of very low safety significance because the floor drain sump pump failure did not increase the likelihood of a plant trip or degrade more than one train of any safety system. The finding is documented in the licensee's corrective action program as CR-RBS-2003-2368.
Inspection Report# : 2003004(pdf)
Significance:      May 07, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Automatic Initiation of Standby Service Water System Due to Inadequate Control of System Operation The inspectors identified a self-revealing violation of Technical Specification 5.4.1 because operators lined up service water to the reactor plant and turbine plant cooling water systems such that an automatic start of standby service water occurred on low system pressure while shifting normal service water pumps. Three heat exchangers in each system were in service when the system operating procedures for reactor plant and turbine plant cooling water allow only two heat exchangers in operation per system.
This finding is greater than minor because it was associated with the ability to meet the mitigating systems cornerstone file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            04/22/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                      Page 3 of 9 objective and because a plant transient occurred. The inspectors determined that the finding was of very low safety significance (Green), since the finding did not represent an actual loss of safety function of a single train.
Inspection Report# : 2003005(pdf)
Significance:      Feb 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain watertight integrity of severe weather doors compromised the availability of standby service water system The inspectors identified a noncited violation for the failure of the licensee to comply with 10 CFR Part 50, Appendix B, Criterion III, "Design Control." This violation was for failure to incorporate necessary measures into station procedures to ensure that the design basis of the doors at the end of underground G-Tunnel was maintained.
The finding was more than minor because it was associated with flood protection measures and degraded the ability to meet the mitigating systems cornerstone objective. It had an adverse impact on the flooding potential of the G-Tunnel, which opened into the base of the standby cooling tower, and challenged the availability of the standby service water system. The finding is of very low safety significance because of the existing condition of the door seals, the availability of two nonsafety-related sump pumps at the base of the standby cooling tower, the relative height of the control circuits and motor operators of the cooling tower inlet valves and the possibility of operator action to manually initiate standby service water before the failure of the standby cooling tower inlet valves. The finding was documented in the licensee's corrective action program as CR-RBS-2003-1894.
Inspection Report# : 2003004(pdf)
Significance:      Nov 14, 2002 Identified By: Self Disclosing Item Type: VIO Violation Failure to properly lock open condensate valve resulted in loss of feedwater flow following reactor scram As documented in special inspection report 05000458/2002007, the inspectors identified a violation of Technical Specifications 5.4.1.a. for failure to properly lock open condensate prefilter vessel bypass flow control Valve CNM-FCV200. As a result, when the reactor automatically scrammed the valve closed and feedwater flow was lost to the reactor. The operators were able to provide makeup water to the reactor using the reactor core isolation cooling system.
The final significance determination was completed and documented in "Final Significance Determination for a White Finding and Notice of Violation," (EA-03-077) dated December 29, 2003. The finding was determined to be of low to moderate safety significance because of the combination of risk associated with a loss of feedwater and from external events, such as a fire in conjunction with a loss of the feedwater system, over a period of approximately 126 days.
Inspection Report# : 2003006(pdf)
Barrier Integrity Significance:      Nov 07, 2003 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                    Page 4 of 9 Failure to Identify Failed Open Secondary Containment Doors as Condition Adverse to Quality The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to identify conditions that would have caused unexpected entry into Technical Specification Action Statements and had the potential to cause secondary containment to be inoperable.
The issue was more than minor because it affects the reactor safety/barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events. The results of the phase one evaluation of the significance determination process was that the issue was of very low safety significance because the finding only represents a degradation of the radiological barrier function provided by the auxiliary building and the duration of each of the 9 incidents was less than 10 minutes.
Inspection Report# : 2003007(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Aug 22, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses associated with RWP 2003-1800 ALARA The inspector identified an ALARA finding because performance deficiencies resulted in a collective dose of the work activity that exceeded 5 person-rem and exceeded the legitimate dose estimation by more than 50 percent. Specifically, Radiation Work Permit 2003-1800, "RF-11 Refueling Activities," accrued 34.962 person-rem and exceeded the dose estimate (19.939 person-rem) by 75 percent. A primary cause for the unplanned dose was the licensee's failure to effectively schedule the use of the Alternate Heat Decay Removal System, a system which had previously proven to be effective at removing radioactivity from the refueling pool. The licensee also failed to limit the number of personnel on the refueling bridge to the planned number, thus causing the work activity to accrue more collective dose than estimated. A contamination incident during the disassembly of the reactor vessel was caused by poor planning and required additional time for cleanup.
This finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/estimated dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). The finding involved a failure to maintain or implement, to the extent practical, procedures or engineering controls needed to achieve occupational doses that were ALARA, and that resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this ALARA finding was found to have no more than very low safety significance because the licensee's 3-year rolling average collective dose was not greater than 240 person-rem.
Inspection Report# : 2003005(pdf)
Significance:      Aug 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                        04/22/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                    Page 5 of 9 Failure to post a radiological hazard The team identified a non-cited violation of Technical Specification 5.4.1a because the licensee failed to post a radiological hazard (hot spot). Station Procedure RP-109, "Hot Spot Program," Revision 0, Step 5.2.1, required that hot spots are identified with a hot spot tag to alert workers of the hazard. However, on August 19, 2003, the team identified a hot spot on an accessible drain line from the radwaste sample sink reading 200 millirem per hour on contact and 50 millirem per hour at one foot from the source. The licensee performed a survey 11 days earlier that identified the radiation levels, however, the technician and the survey reviewer failed to tag the hot spot to warn workers of the hazard.
The finding was more than minor because it was associated with the Occupational Radiation Safety cornerstone attribute (Program and Process) and affected the associated cornerstone objective. The finding involved the potential for a workers unplanned or unintended dose resulting from actions contrary to procedures. When processed through the Occupational Radiation Safety Significance Determination Process the team determined that the finding had very low safety significance because the finding did not involve as low as is reasonably achievable (ALARA) planning or work controls, no individual received an overexposure or a substantial potential for overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2003009(pdf)
Significance:      Apr 09, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses ALARA that were associated with RWP 2003-1929 A finding was identified because the licensee failed to maintain collective doses ALARA. Specifically, the work activity collective dose associated with RWP 2003-1929 Task 1, "Refueling Outage 11 Recirculation Pump Work,"
exceeded 5 person-rem and exceeded the dose estimation by more than 50 percent.
The failure to maintain collective doses ALARA is a performance deficiency. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/projected dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). This occurrence involved worker inefficiencies, inadequate planning, scheduling and supervisory oversight which resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this finding was found to have no more than very low safety significance because the finding was an ALARA Planning issue, the licensee's three-year rolling average collective dose was greater than 240 person-rem, the actual dose for the work activity was not more than 25 person-rem, and there were no more than four occurrences.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses ALARA that were associated with RWP 2003-1935 A finding was identified because the licensee failed to maintain collective doses ALARA. Specifically, the work activity collective dose associated with RWP 2003-1935, "Drywell Valve Maintenance, to include Repacks and Support Work," exceeded 5 person-rem and exceeded the dose estimation by more than 50 percent.
The failure to maintain collective doses ALARA is a performance deficiency. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/projected dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                        04/22/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                    Page 6 of 9 exposure to radiation). This occurrence involved worker inefficiencies and inadequate planning which resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this finding was found to have no more than very low safety significance because the finding was an ALARA Planning issue, the licensee's three-year rolling average collective dose was greater than 240 person-rem, the actual dose for the work activity was not more than 25 person-rem, and there were no more than four occurrences.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to develop a sufficiently detailed work plan The licensee failed to develop a sufficiently detailed work plan for the decontamination of the reactor vessel bellows, in violation of Technical Specification 5.4.1. a. The work plan failed to provide guidance on maintaining highly contaminated surfaces (the reactor vessel bellows surface) wet, using a hydrolaser with a rotary surface cleaner, or briefing the individual using the hydrolaser. The lack of a detailed work planned contributed to an unexpected increase in airborne radioactivity and unplanned personnel exposures.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an as low as is reasonably achievable (ALARA) finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to survey The licensee failed, in three examples, to survey or evaluate radiological hazards when conducting reactor vessel bellows decontamination, in violation of 10 CFR 20.1501(a). First, the licensee failed to evaluate the highest concentration of radioactive contamination on the reactor vessel bellows. Additionally, the licensee failed to evaluate the airborne radioactivity in the immediate vicinity of reactor vessel bellows contamination. Later, the licensee failed to evaluate the airborne radioactivity levels throughout the containment building when continuous air monitors alarmed.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, there was no personnel overexposure, there was no substantial potential for personnel overexposure, and the finding did not file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                        04/22/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                      Page 7 of 9 compromise the licensee's ability to assess dose, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to post an airborne radioactivity area The licensee failed to post the reactor containment building as an airborne radioactivity area, in violation with 10 CFR 20.1902(d). Airborne radioactivity levels exceeded the allowable limits in 10 CFR Part 20, Appendix B by as much as 3.5 times. The condition existed for at least five hours.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to instruct workers Following an occurrence that caused an airborne radioactivity area, the licensee failed to inform workers of radiological conditions that had changed and of precautions to minimize exposure, in violation of 10 CFR 19.12.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:      Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to control a locked high radiation area The licensee failed to control an area with dose rates of 1000 millirems per hour as a locked high radiation area, in violation of Technical Specification 5.7.2. After a plant scram on September 18, 2002, a worker entered the reactor core isolation cooling area on the 95-foot elevation of the auxiliary building and received an electronic dosimeter dose file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                      Page 8 of 9 rate alarm. A crud burst resulting from a transient that occurred approximately three hours previously caused the dose levels in the area entered by the worker to increase to 1000 millirems per hour. Historically, the site has experienced crud bursts under similar conditions and the increase in dose rate should have been anticipated and evaluated.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Public Radiation Safety Significance:      Aug 22, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to properly control radioactive material A self-revealing non-cited violation of Technical Specification 5.4.1a was reviewed by the team because the licensee did not prevent the release of detectable licensed radioactive material from the controlled access area. Specifically, Section 5.1.1 of Procedure RSP-213, "Control and Handling of Radioactive Materials," Revision 16, stated, in part, that material can be unconditionally released from the controlled access area if there is no detectable loose surface and fixed contamination above background radiation levels. However, on March 31, 2003, the licensee failed to evaluate an item, against their procedural criteria, prior to it being unconditionally released from the controlled access area and subsequently released from the protected area. Fixed contamination levels were as high as 1,000 corrected counts per minute per probe area.
The finding was more than minor because it was associated with the Public Radiation Safety cornerstone attribute (Program and Process) and affected the associated cornerstone objective. The finding involved an occurrence in the radioactive material control program that was contrary to licensee procedures. When processed through the Public Radiation Safety Significance Determination Process, the team determined the finding had very low safety significance because the public exposure associated with the item was less than 5 millirem and there were not more than 5 occurrences.
Inspection Report# : 2003009(pdf)
Physical Protection Significance: N/A Aug 07, 2003 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - River Bend 1                                                                        Page 9 of 9 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 Nov 07, 2003 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Program Assessment The team concluded that the licensee was effective at identifying problems and putting them into the corrective action program. The licensee's effectiveness at problem identification was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee, during the review period. However, the team identified a repetitive failure on the part of the licensee to properly identify the inability of secondary containment doors to close and potential failures of secondary containment. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementing corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. On the basis of interviews conducted during this inspection, workers at the site felt free to input safety findings into the problem identification and resolution program.
Inspection Report# : 2003007(pdf)
Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\RBS1\rbs1_pim.html                                                            04/22/2004
 
1Q/2004 Inspection Findings - River Bend 1                                                                                            Page 1 of 7 River Bend 1 1Q/2004 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding Normal Service Water pump found to be air-bound when called upon to run A self-revealing finding was identified and determined to be of very low safety significance. A human performance error caused the isolation of the air release valve for normal service water Pump C. The air release valve for a normal service water pump served as a high point vent on the system while the pump was secured. As a result, normal service water Pump C became air bound while in standby, and failed to develop discharge pressure when started during a manual swap of running normal service water pumps on September 1, 2003. The inspectors determined that the finding did not represent a noncompliance because it occurred on non-safety-related normal service water system.
The inspectors determined that the failure to maintain normal service water Pump C discharge air release valve isolation valve open was more than minor because it was associated with an increase in the likelihood of an initiating event. The finding was of very low safety significance because there was only a small increase in the likelihood of a loss of normal service water with one of the three 50 percent capacity normal service water pumps unavailable and because the standby service water system was available throughout the time normal service Pump C was air bound.
Inspection Report# : 2003006(pdf)
Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding Repeat failure of main turbine control system hydraulic lines leads to manual reactor scram and turbine trip A self-revealing finding was identified for failure to properly diagnose a failure of turbine control hydraulic line failure in August 2000. As a result, a similar line failed on February 22, 2003, causing the operators to scram the reactor and trip the main turbine. The inspectors determined that the finding did not represent a noncompliance because it occurred on non-safety-related secondary plant equipment.
The inspectors determined that this problem identification and resolution finding is more than minor because the misdiagnosis of the August 31, 2000 failure contributed to the February 22, 2003, failure. The finding affected the initiating events cornerstone and was considered to have very low safety significance because it did not contribute to the likelihood of a LOCA, nor the likelihood of both a reactor scram and mitigating equipment or functions being unavailable, and because there was no increased likelihood of a fire or internal/external flood.
Inspection Report# : 2003006(pdf)
Mitigating Systems Significance:        Feb 13, 2004 Identified By: Licensee Item Type: FIN Finding Untimely Corrective Actions for Degraded Fire Protection Feature The licensee relied on compensatory measures for seven years instead of correcting a fire protection coating deficiency in three areas important to safe shutdown. In 1997, the licensee identified that the fire protective coatings on most structural steel beams in safety-related buildings did not meet the required thickness for a 3-hour fire rating. The deficient condition typically existed over one-fourth of each beam. While the majority of the deficiencies were repaired by building up the thickness, three fire areas remain degraded and had been subject to hourly fire watches since 1997. The team concluded that the planned corrective actions to restore the fire protection feature to its required condition for the remaining degraded areas were not timely.
This finding was greater than minor because it was similar to example 2.e in Appendix E of Manual Chapter 0609 and the finding is associated with degradation of a fire protection feature. This finding screened as having very low safety significance because the compensatory fire watches were in place as required and the remaining defense in depth elements remained unaffected.
Inspection Report# : 2004007(pdf) 07/14/2004
 
1Q/2004 Inspection Findings - River Bend 1                                                                                            Page 2 of 7 Significance:        Feb 13, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Ventilation and Smoke Control Associated With a Fire The inspectors identified a non-cited violation of License Condition 2.C(10) and by reference the fire protection program and Appendix R to 10 CFR 50, Section III.K.12.h. The non-cited violation was identified related to fire response procedures and pre-fire strategies that did not contain adequate procedure steps for controlling the ventilation system alignment in order to both remove smoke and assure adequate cooling to remaining safe shutdown equipment. The team identified that the licensee did not account for fire dampers with heat-activated fusible links throughout the system, which could reasonably be expected to close when hot smoke was passed through the dampers. The licensee made a prompt change to FPP-0010, "Fire Fighting Procedure," to make operators aware of the condition as a compensatory measure. This issue was entered into the licensee's corrective action program under Condition Report 2004-000276.
This finding was greater than minor because it affected the Mitigating Systems Cornerstone objective of equipment reliability, in that loss of cooling or exposure to smoke and hot gases could cause failure of safe shutdown equipment that was supposed to remain unaffected by a particular fire. This finding screened as having very low safety significance because it affects a fire protection feature that was not a defense in depth element.
Inspection Report# : 2004007(pdf)
Significance:        Jun 17, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation HPCS Inadvertently Disabled Due to Personnel Error During Installation of Clearance Order The inspectors identified a self-revealing violation for failure to comply with Technical Specification 5.4.1.a. Operators mistakenly racked out the HPCS pump breaker when implementing a clearance order on a standby service water.
This self-revealing finding was more than minor because the HPCS safety function was made unavailable. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations."
The inspectors determined that the finding was of very low safety significance (Green) because the HPCS pump was not functional for less than one hour. Recovery credit was given for operator actions necessary to restore the equipment lineup and recover the safety function.
Inspection Report# : 2003005(pdf)
Significance:        Jun 10, 2003 Identified By: Self Disclosing Item Type: FIN Finding Foreign material caused failure of one residual heat removal equipment room floor drain sump pump while the other pump was unavailable The inspectors identified a self-revealing finding for failure to control foreign material in the residual heat removal Train B equipment room which resulted in the failure of one of two floor drain pumps while the other floor drain pump was unavailable. The finding was of very low safety significance because the floor drain sump pump failure did not cause an actual loss of safety function for residual heat removal Train B.
The inspectors determined that the licensee's failure to control foreign material in the residual heat removal Train B equipment room, which resulted in the fouling and unavailability of floor drain Pump DFR-P3L while Pump DFR-P3E was also unavailable, was a performance deficiency. This self-revealing finding was more than minor because, if left uncorrected and a leak developed in the residual heat removal Train B equipment room, the unavailability of both floor drain sump pumps could lead to a loss of residual heat removal Train B. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the phase one screening of the finding, the inspectors determined that the finding was of very low safety significance because the floor drain sump pump failure did not increase the likelihood of a plant trip or degrade more than one train of any safety system. The finding is documented in the licensee's corrective action program as CR-RBS-2003-2368.
Inspection Report# : 2003004(pdf)
Significance:        May 07, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Automatic Initiation of Standby Service Water System Due to Inadequate Control of System Operation The inspectors identified a self-revealing violation of Technical Specification 5.4.1 because operators lined up service water to the reactor plant and turbine plant cooling water systems such that an automatic start of standby service water occurred on low system pressure while shifting normal service water pumps. Three heat exchangers in each system were in service when the system operating procedures for reactor plant and turbine plant cooling water allow only two heat exchangers in operation per system.
07/14/2004
 
1Q/2004 Inspection Findings - River Bend 1                                                                                          Page 3 of 7 This finding is greater than minor because it was associated with the ability to meet the mitigating systems cornerstone objective and because a plant transient occurred. The inspectors determined that the finding was of very low safety significance (Green), since the finding did not represent an actual loss of safety function of a single train.
Inspection Report# : 2003005(pdf)
Barrier Integrity Significance:        Nov 07, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Failed Open Secondary Containment Doors as Condition Adverse to Quality The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to identify conditions that would have caused unexpected entry into Technical Specification Action Statements and had the potential to cause secondary containment to be inoperable.
The issue was more than minor because it affects the reactor safety/barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events. The results of the phase one evaluation of the significance determination process was that the issue was of very low safety significance because the finding only represents a degradation of the radiological barrier function provided by the auxiliary building and the duration of each of the 9 incidents was less than 10 minutes.
Inspection Report# : 2003007(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Aug 22, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses associated with RWP 2003-1800 ALARA The inspector identified an ALARA finding because performance deficiencies resulted in a collective dose of the work activity that exceeded 5 person-rem and exceeded the legitimate dose estimation by more than 50 percent. Specifically, Radiation Work Permit 2003-1800, "RF-11 Refueling Activities," accrued 34.962 person-rem and exceeded the dose estimate (19.939 person-rem) by 75 percent. A primary cause for the unplanned dose was the licensee's failure to effectively schedule the use of the Alternate Heat Decay Removal System, a system which had previously proven to be effective at removing radioactivity from the refueling pool. The licensee also failed to limit the number of personnel on the refueling bridge to the planned number, thus causing the work activity to accrue more collective dose than estimated. A contamination incident during the disassembly of the reactor vessel was caused by poor planning and required additional time for cleanup.
This finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/estimated dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). The finding involved a failure to maintain or implement, to the extent practical, procedures or engineering controls needed to achieve occupational doses that were ALARA, and that resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this ALARA finding was found to have no more than very low safety significance because the licensee's 3-year rolling average collective dose was not greater than 240 person-rem.
Inspection Report# : 2003005(pdf)
Significance:        Aug 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a radiological hazard The team identified a non-cited violation of Technical Specification 5.4.1a because the licensee failed to post a radiological hazard (hot spot).
07/14/2004
 
1Q/2004 Inspection Findings - River Bend 1                                                                                            Page 4 of 7 Station Procedure RP-109, "Hot Spot Program," Revision 0, Step 5.2.1, required that hot spots are identified with a hot spot tag to alert workers of the hazard. However, on August 19, 2003, the team identified a hot spot on an accessible drain line from the radwaste sample sink reading 200 millirem per hour on contact and 50 millirem per hour at one foot from the source. The licensee performed a survey 11 days earlier that identified the radiation levels, however, the technician and the survey reviewer failed to tag the hot spot to warn workers of the hazard.
The finding was more than minor because it was associated with the Occupational Radiation Safety cornerstone attribute (Program and Process) and affected the associated cornerstone objective. The finding involved the potential for a workers unplanned or unintended dose resulting from actions contrary to procedures. When processed through the Occupational Radiation Safety Significance Determination Process the team determined that the finding had very low safety significance because the finding did not involve as low as is reasonably achievable (ALARA) planning or work controls, no individual received an overexposure or a substantial potential for overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2003009(pdf)
Significance:        Apr 09, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses ALARA that were associated with RWP 2003-1929 A finding was identified because the licensee failed to maintain collective doses ALARA. Specifically, the work activity collective dose associated with RWP 2003-1929 Task 1, "Refueling Outage 11 Recirculation Pump Work," exceeded 5 person-rem and exceeded the dose estimation by more than 50 percent.
The failure to maintain collective doses ALARA is a performance deficiency. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/projected dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). This occurrence involved worker inefficiencies, inadequate planning, scheduling and supervisory oversight which resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this finding was found to have no more than very low safety significance because the finding was an ALARA Planning issue, the licensee's three-year rolling average collective dose was greater than 240 person-rem, the actual dose for the work activity was not more than 25 person-rem, and there were no more than four occurrences.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses ALARA that were associated with RWP 2003-1935 A finding was identified because the licensee failed to maintain collective doses ALARA. Specifically, the work activity collective dose associated with RWP 2003-1935, "Drywell Valve Maintenance, to include Repacks and Support Work," exceeded 5 person-rem and exceeded the dose estimation by more than 50 percent.
The failure to maintain collective doses ALARA is a performance deficiency. This finding was more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/projected dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). This occurrence involved worker inefficiencies and inadequate planning which resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this finding was found to have no more than very low safety significance because the finding was an ALARA Planning issue, the licensee's three-year rolling average collective dose was greater than 240 person-rem, the actual dose for the work activity was not more than 25 person-rem, and there were no more than four occurrences.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to develop a sufficiently detailed work plan The licensee failed to develop a sufficiently detailed work plan for the decontamination of the reactor vessel bellows, in violation of Technical Specification 5.4.1. a. The work plan failed to provide guidance on maintaining highly contaminated surfaces (the reactor vessel bellows surface) wet, using a hydrolaser with a rotary surface cleaner, or briefing the individual using the hydrolaser. The lack of a detailed work planned contributed to an unexpected increase in airborne radioactivity and unplanned personnel exposures.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from 07/14/2004
 
1Q/2004 Inspection Findings - River Bend 1                                                                                            Page 5 of 7 actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an as low as is reasonably achievable (ALARA) finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to survey The licensee failed, in three examples, to survey or evaluate radiological hazards when conducting reactor vessel bellows decontamination, in violation of 10 CFR 20.1501(a). First, the licensee failed to evaluate the highest concentration of radioactive contamination on the reactor vessel bellows. Additionally, the licensee failed to evaluate the airborne radioactivity in the immediate vicinity of reactor vessel bellows contamination. Later, the licensee failed to evaluate the airborne radioactivity levels throughout the containment building when continuous air monitors alarmed.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to post an airborne radioactivity area The licensee failed to post the reactor containment building as an airborne radioactivity area, in violation with 10 CFR 20.1902(d). Airborne radioactivity levels exceeded the allowable limits in 10 CFR Part 20, Appendix B by as much as 3.5 times. The condition existed for at least five hours.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to instruct workers Following an occurrence that caused an airborne radioactivity area, the licensee failed to inform workers of radiological conditions that had changed and of precautions to minimize exposure, in violation of 10 CFR 19.12.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf) 07/14/2004
 
1Q/2004 Inspection Findings - River Bend 1                                                                                              Page 6 of 7 Significance:        Apr 09, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to control a locked high radiation area The licensee failed to control an area with dose rates of 1000 millirems per hour as a locked high radiation area, in violation of Technical Specification 5.7.2. After a plant scram on September 18, 2002, a worker entered the reactor core isolation cooling area on the 95-foot elevation of the auxiliary building and received an electronic dosimeter dose rate alarm. A crud burst resulting from a transient that occurred approximately three hours previously caused the dose levels in the area entered by the worker to increase to 1000 millirems per hour.
Historically, the site has experienced crud bursts under similar conditions and the increase in dose rate should have been anticipated and evaluated.
This self-revealing, noncited violation was greater than minor because it was associated with one of the Occupational Radiation Safety Cornerstone attributes (exposure/contamination control) and the finding affected the associated cornerstone objective (to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material). The inspector processed the violation through the Occupational Radiation Protection Significance Determination Process because the occurrence involved potential doses (resulting from actions or conditions contrary to licensee procedures) which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances. However, because the violation was not an ALARA finding, 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, the violation had no more than very low safety significance.
Inspection Report# : 2003003(pdf)
Public Radiation Safety Significance:        Aug 22, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to properly control radioactive material A self-revealing non-cited violation of Technical Specification 5.4.1a was reviewed by the team because the licensee did not prevent the release of detectable licensed radioactive material from the controlled access area. Specifically, Section 5.1.1 of Procedure RSP-213, "Control and Handling of Radioactive Materials," Revision 16, stated, in part, that material can be unconditionally released from the controlled access area if there is no detectable loose surface and fixed contamination above background radiation levels. However, on March 31, 2003, the licensee failed to evaluate an item, against their procedural criteria, prior to it being unconditionally released from the controlled access area and subsequently released from the protected area. Fixed contamination levels were as high as 1,000 corrected counts per minute per probe area.
The finding was more than minor because it was associated with the Public Radiation Safety cornerstone attribute (Program and Process) and affected the associated cornerstone objective. The finding involved an occurrence in the radioactive material control program that was contrary to licensee procedures. When processed through the Public Radiation Safety Significance Determination Process, the team determined the finding had very low safety significance because the public exposure associated with the item was less than 5 millirem and there were not more than 5 occurrences.
Inspection Report# : 2003009(pdf)
Physical Protection Significance: N/A Aug 07, 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) 07/14/2004
 
1Q/2004 Inspection Findings - River Bend 1                                                                                            Page 7 of 7 Miscellaneous Significance: N/A Nov 07, 2003 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Program Assessment The team concluded that the licensee was effective at identifying problems and putting them into the corrective action program. The licensee's effectiveness at problem identification was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee, during the review period. However, the team identified a repetitive failure on the part of the licensee to properly identify the inability of secondary containment doors to close and potential failures of secondary containment.
The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementing corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. On the basis of interviews conducted during this inspection, workers at the site felt free to input safety findings into the problem identification and resolution program.
Inspection Report# : 2003007(pdf)
Last modified : May 05, 2004 07/14/2004
 
2Q/2004 Inspection Findings - River Bend 1                                                                                                        Page 1 of 5 River Bend 1 2Q/2004 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding Normal Service Water pump found to be air-bound when called upon to run A self-revealing finding was identified and determined to be of very low safety significance. A human performance error caused the isolation of the air release valve for normal service water Pump C. The air release valve for a normal service water pump served as a high point vent on the system while the pump was secured. As a result, normal service water Pump C became air bound while in standby, and failed to develop discharge pressure when started during a manual swap of running normal service water pumps on September 1, 2003. The inspectors determined that the finding did not represent a noncompliance because it occurred on non-safety-related normal service water system.
The inspectors determined that the failure to maintain normal service water Pump C discharge air release valve isolation valve open was more than minor because it was associated with an increase in the likelihood of an initiating event. The finding was of very low safety significance because there was only a small increase in the likelihood of a loss of normal service water with one of the three 50 percent capacity normal service water pumps unavailable and because the standby service water system was available throughout the time normal service Pump C was air bound.
Inspection Report# : 2003006(pdf)
Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding Repeat failure of main turbine control system hydraulic lines leads to manual reactor scram and turbine trip A self-revealing finding was identified for failure to properly diagnose a failure of turbine control hydraulic line failure in August 2000. As a result, a similar line failed on February 22, 2003, causing the operators to scram the reactor and trip the main turbine. The inspectors determined that the finding did not represent a noncompliance because it occurred on non-safety-related secondary plant equipment.
The inspectors determined that this problem identification and resolution finding is more than minor because the misdiagnosis of the August 31, 2000 failure contributed to the February 22, 2003, failure. The finding affected the initiating events cornerstone and was considered to have very low safety significance because it did not contribute to the likelihood of a LOCA, nor the likelihood of both a reactor scram and mitigating equipment or functions being unavailable, and because there was no increased likelihood of a fire or internal/external flood.
Inspection Report# : 2003006(pdf)
Mitigating Systems Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control conditions of engineered safety features electrical switchgear.
The inspectors identified two examples of a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to maintain the original design configuration of engineered safety feature switchgear. The inspectors found all of the heat dissipation louvers on top of the load centers and the relay control cabinets for both Divisions I and II auxiliary building 480 Vac engineered safety features switchgear covered with tape. Previously, the licensee had identified cardboard covering the ventilation louvers on breaker cubicles in the Division I engineered safety features 4160 Vac switchgear in the control building.
The failure to maintain design control over Switchgear EJS-SWGR2A and -2B and ENS-SWGR1A was a performance deficiency. The violation was more than minor because it was associated with the mitigating systems cornerstone attribute for design control. It affects the mitigating system cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. This noncited violation was evaluated using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." During the Phase 1 analysis, the issue was determined to have very low safety significance because it did not: (1) represent a design or qualification deficiency, (2) represent an actual loss of safety function of a system or a single train of a system for greater than the Technical Specification allowed out-of-service time, (3) represent an actual loss of safety function of non-Technical Specification trains of equipment per 10 CFR 50.65 for more than 24 hours, and (4) screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Since this violation of 10 CFR Part 50, Appendix B, Criterion III, was of very low safety significance and was entered in the licensee's corrective action program as CR-RBS-2004-0512, -1389, -1855, and -1856, it is being treated as a noncited violation consistent with the NRC Enforcement Policy, NUREG-1600.
 
2Q/2004 Inspection Findings - River Bend 1                                                                                                      Page 2 of 5 The inspectors also determined that on at least two occasions the licensee had the opportunity but failed to identify the tape covering the louvers on top of auxiliary building 480 Vac engineered safety features Switchgear EJS-SWGR2A. Therefore, the inspectors consider this finding to have problem identification and resolution crosscutting aspects for failure to identify a condition adverse to quality. Also the inspectors determined that the design engineering evaluation of as-found conditions for Division I engineered safety features 4160 Vac ENS-SWGR1A for past reportability was actually an evaluation of Division I 480 Vac engineered safety features EJS-SWGR1A and therefore a human performance error.
Inspection Report# : 2004003(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable preconditioning of Technical Specification diesel generator surveillance testing.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's performance of unacceptable preconditioning of Technical Specification emergency diesel generator surveillance testing. The inspectors found three unacceptable preconditioning activities the licensee performed during the May and June 2004 emergency diesel generator monthly surveillance tests.
The inspectors determined that this finding has problem identification and resolution aspects because the licensee identified some of these activities as unacceptable preconditioning in their evaluation of NRC Information Notice 97-16, "Preconditioning of Plant Structures, Systems, and Components Before ASME Code Inservice Testing or Technical Specification Surveillance Testing," dated June 9, 1997, yet failed to take actions to correct the test procedures.
The inspectors determined the unacceptable preconditioning of emergency diesel generator surveillance testing was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute for procedure quality. The finding affected the cornerstone objective to maintain availability and reliability of systems that respond to events to prevent undesirable consequences. The inspectors reviewed the finding using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations."
Based on the results of the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, was not an actual loss of safety function for a system or train, and was not risk significant due to a seismic, fire, flooding, or sever weather initiating event. The inspectors determined that unacceptable preconditioning of Technical Specification diesel generator surveillance testing was a violation of 10 CFR Part 50, Appendix B, Criterion V. Because the violation was of very low safety significance and was entered into the licensee's corrective action program as CR-RBS-2004-1839 and -1858, it is being treated as an noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG 1600.
The inspectors identified crosscutting aspects related to problem identification and resolution. In their evaluation of NRC Information Notice 97-15, the licensee identified and evaluated some activities that precondition the emergency diesel generators during their prestart checks for surveillance testing, but failed to take appropriate actions to correct the procedures.
Inspection Report# : 2004003(pdf)
Significance:        Apr 06, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to adequately address roof leaks in the auxiliary building resulted in electrical grounds on safety related switchgear The licensee failed to adequately address leaks in the roof of the auxiliary building following several instances when roof leaks were identified and documented in the licensee's corrective action program. On February 5, 2004, rainwater inleakage through the auxiliary building roof resulted in an electrical ground on the control circuits of auxiliary building 480 Vac engineered safety features Switchgear EJS-SWG2A. The finding was of very low safety significance because, although it degraded one train of safety-related equipment, and could have degraded it again, it did not: increase the likelihood of a primary or secondary system loss of coolant accident initiator, contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, or increase the likelihood of a fire or internal/external flood.
The inspectors determined that the failure to correct the leaks in the auxiliary building was a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." Because this problem identification and resolution finding was of very low safety significance and was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-01083, it is being treated as an noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-16000.
Inspection Report# : 2004002(pdf)
Significance:        Feb 13, 2004 Identified By: Licensee Item Type: FIN Finding Untimely Corrective Actions for Degraded Fire Protection Feature The licensee relied on compensatory measures for seven years instead of correcting a fire protection coating deficiency in three areas important to safe shutdown. In 1997, the licensee identified that the fire protective coatings on most structural steel beams in safety-related buildings did not meet the required thickness for a 3-hour fire rating. The deficient condition typically existed over one-fourth of each beam. While the majority of the deficiencies were repaired by building up the thickness, three fire areas remain degraded and had been subject to hourly fire watches since 1997. The team concluded that the planned corrective actions to restore the fire protection feature to its required condition for the remaining degraded areas were not timely.
This finding was greater than minor because it was similar to example 2.e in Appendix E of Manual Chapter 0609 and the finding is associated with degradation of a fire protection feature. This finding screened as having very low safety significance because the compensatory fire watches were in
 
2Q/2004 Inspection Findings - River Bend 1                                                                                                      Page 3 of 5 place as required and the remaining defense in depth elements remained unaffected.
Inspection Report# : 2004007(pdf)
Significance:        Feb 13, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Ventilation and Smoke Control Associated With a Fire The inspectors identified a non-cited violation of License Condition 2.C(10) and by reference the fire protection program and Appendix R to 10 CFR 50, Section III.K.12.h. The non-cited violation was identified related to fire response procedures and pre-fire strategies that did not contain adequate procedure steps for controlling the ventilation system alignment in order to both remove smoke and assure adequate cooling to remaining safe shutdown equipment. The team identified that the licensee did not account for fire dampers with heat-activated fusible links throughout the system, which could reasonably be expected to close when hot smoke was passed through the dampers. The licensee made a prompt change to FPP-0010, "Fire Fighting Procedure," to make operators aware of the condition as a compensatory measure. This issue was entered into the licensee's corrective action program under Condition Report 2004-000276.
This finding was greater than minor because it affected the Mitigating Systems Cornerstone objective of equipment reliability, in that loss of cooling or exposure to smoke and hot gases could cause failure of safe shutdown equipment that was supposed to remain unaffected by a particular fire. This finding screened as having very low safety significance because it affects a fire protection feature that was not a defense in depth element.
Inspection Report# : 2004007(pdf)
Barrier Integrity Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a functional failure.
The NRC inspectors identified a noncited violation of 10 CFR 50.65(a)(2). On May 15, 2003, the licensee failed to set goals and monitor the performance of the secondary containment system as required by 10 CFR 50.65(a)(1). As required by 10 CFR 50.65(a)(2), the licensee must demonstrate effective control of a structure's condition through appropriate preventive maintenance to not require paragraph (a)(1) monitoring. The licensee had no justification for not requiring (a)(1) monitoring, after they failed to demonstrate effective control of the performance of the secondary containment system through appropriate preventive maintenance. The inspectors considered this violation to be noncited consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee entered this noncited violation into its corrective action program as Condition Report CR-RBS-2004-01706.
The inspectors determined this violation was more than minor because the failure to identify functional failures resulted in the system not being evaluated for 10 CFR 50.65(a)(1) status and had a credible impact on safety. The licensee performed engineering evaluations which concluded that, had a design basis accident occurred while the condition existed, the main control room, exclusion area boundary, and low population zone doses would have remained within the limits of 10CFR50.67. The inspectors determined the safety significance of this violation to be very low by the Reactor Safety Significance Determination Process. The inspectors answered the Phase 1 question regarding containment as yes because the inspectors determined that this finding represented a degradation of the radiological barrier only; therefore, in accordance with Manual Chapter 0609, Appendix A, Attachment 1, this finding is of very low safety significance.
Inspection Report# : 2004003(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Failed Open Secondary Containment Doors as Condition Adverse to Quality The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to identify conditions that would have caused unexpected entry into Technical Specification Action Statements and had the potential to cause secondary containment to be inoperable.
The issue was more than minor because it affects the reactor safety/barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events. The results of the phase one evaluation of the significance determination process was that the issue was of very low safety significance because the finding only represents a degradation of the radiological barrier function provided by the auxiliary building and the duration of each of the 9 incidents was less than 10 minutes.
Inspection Report# : 2003007(pdf)
Emergency Preparedness
 
2Q/2004 Inspection Findings - River Bend 1                                                                                                      Page 4 of 5 Occupational Radiation Safety Significance:        Mar 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify the correct configuration and adequacy of permanent shielding The inspectors identified a non-cited violation of Technical Specification 5.4.1.a because the licensee failed to follow procedural requirements to verify the correct configuration and adequacy of permanent shielding. On March 25, 2004, the inspectors identified that permanent shielding on a low-pressure core spray flush line, in the crescent area of the 70-foot elevation of the Auxiliary Building, was not in the correct configuration and not adequate for the intended application.
The failure to verify the correct configuration of permanent shielding and ensure that it was adequate for the intended application was 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. When processed through the Occupational Radiation Safety Significance Determination Process the finding was determined to be of very low safety significance because the finding was not associated with as low as is reasonably achievable issues, there was no overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. The finding was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-00924.
Inspection Report# : 2004002(pdf)
Significance:        Aug 22, 2003 Identified By: NRC Item Type: FIN Finding Failure to maintain collective doses associated with RWP 2003-1800 ALARA The inspector identified an ALARA finding because performance deficiencies resulted in a collective dose of the work activity that exceeded 5 person-rem and exceeded the legitimate dose estimation by more than 50 percent. Specifically, Radiation Work Permit 2003-1800, "RF-11 Refueling Activities," accrued 34.962 person-rem and exceeded the dose estimate (19.939 person-rem) by 75 percent. A primary cause for the unplanned dose was the licensee's failure to effectively schedule the use of the Alternate Heat Decay Removal System, a system which had previously proven to be effective at removing radioactivity from the refueling pool. The licensee also failed to limit the number of personnel on the refueling bridge to the planned number, thus causing the work activity to accrue more collective dose than estimated. A contamination incident during the disassembly of the reactor vessel was caused by poor planning and required additional time for cleanup.
This finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (ALARA planning/estimated dose) and affected the associated cornerstone objective (to ensure adequate protection of worker health and safety from exposure to radiation). The finding involved a failure to maintain or implement, to the extent practical, procedures or engineering controls needed to achieve occupational doses that were ALARA, and that resulted in unplanned, unintended occupational collective dose for a work activity. When processed through the Occupational Radiation Safety Significance Determination Process, this ALARA finding was found to have no more than very low safety significance because the licensee's 3-year rolling average collective dose was not greater than 240 person-rem.
Inspection Report# : 2003005(pdf)
Significance:        Aug 22, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to post a radiological hazard The team identified a non-cited violation of Technical Specification 5.4.1a because the licensee failed to post a radiological hazard (hot spot). Station Procedure RP-109, "Hot Spot Program," Revision 0, Step 5.2.1, required that hot spots are identified with a hot spot tag to alert workers of the hazard.
However, on August 19, 2003, the team identified a hot spot on an accessible drain line from the radwaste sample sink reading 200 millirem per hour on contact and 50 millirem per hour at one foot from the source. The licensee performed a survey 11 days earlier that identified the radiation levels, however, the technician and the survey reviewer failed to tag the hot spot to warn workers of the hazard.
The finding was more than minor because it was associated with the Occupational Radiation Safety cornerstone attribute (Program and Process) and affected the associated cornerstone objective. The finding involved the potential for a workers unplanned or unintended dose resulting from actions contrary to procedures. When processed through the Occupational Radiation Safety Significance Determination Process the team determined that the finding had very low safety significance because the finding did not involve as low as is reasonably achievable (ALARA) planning or work controls, no individual received an overexposure or a substantial potential for overexposure, and the ability to assess dose was not compromised.
Inspection Report# : 2003009(pdf)
Public Radiation Safety
 
2Q/2004 Inspection Findings - River Bend 1                                                                                                      Page 5 of 5 Significance:        Aug 22, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to properly control radioactive material A self-revealing non-cited violation of Technical Specification 5.4.1a was reviewed by the team because the licensee did not prevent the release of detectable licensed radioactive material from the controlled access area. Specifically, Section 5.1.1 of Procedure RSP-213, "Control and Handling of Radioactive Materials," Revision 16, stated, in part, that material can be unconditionally released from the controlled access area if there is no detectable loose surface and fixed contamination above background radiation levels. However, on March 31, 2003, the licensee failed to evaluate an item, against their procedural criteria, prior to it being unconditionally released from the controlled access area and subsequently released from the protected area.
Fixed contamination levels were as high as 1,000 corrected counts per minute per probe area.
The finding was more than minor because it was associated with the Public Radiation Safety cornerstone attribute (Program and Process) and affected the associated cornerstone objective. The finding involved an occurrence in the radioactive material control program that was contrary to licensee procedures. When processed through the Public Radiation Safety Significance Determination Process, the team determined the finding had very low safety significance because the public exposure associated with the item was less than 5 millirem and there were not more than 5 occurrences.
Inspection Report# : 2003009(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Nov 07, 2003 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Program Assessment The team concluded that the licensee was effective at identifying problems and putting them into the corrective action program. The licensee's effectiveness at problem identification was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee, during the review period. However, the team identified a repetitive failure on the part of the licensee to properly identify the inability of secondary containment doors to close and potential failures of secondary containment. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementing corrective actions.
Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. On the basis of interviews conducted during this inspection, workers at the site felt free to input safety findings into the problem identification and resolution program.
Inspection Report# : 2003007(pdf)
Last modified : September 08, 2004
 
3Q/2004 Inspection Findings - River Bend 1                                                                                              Page 1 of 5 River Bend 1 3Q/2004 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding Normal Service Water pump found to be air-bound when called upon to run A self-revealing finding was identified and determined to be of very low safety significance. A human performance error caused the isolation of the air release valve for normal service water Pump C. The air release valve for a normal service water pump served as a high point vent on the system while the pump was secured. As a result, normal service water Pump C became air bound while in standby, and failed to develop discharge pressure when started during a manual swap of running normal service water pumps on September 1, 2003. The inspectors determined that the finding did not represent a noncompliance because it occurred on non-safety-related normal service water system.
The inspectors determined that the failure to maintain normal service water Pump C discharge air release valve isolation valve open was more than minor because it was associated with an increase in the likelihood of an initiating event. The finding was of very low safety significance because there was only a small increase in the likelihood of a loss of normal service water with one of the three 50 percent capacity normal service water pumps unavailable and because the standby service water system was available throughout the time normal service Pump C was air bound.
Inspection Report# : 2003006(pdf)
Significance:        Dec 31, 2003 Identified By: Self Disclosing Item Type: FIN Finding Repeat failure of main turbine control system hydraulic lines leads to manual reactor scram and turbine trip A self-revealing finding was identified for failure to properly diagnose a failure of turbine control hydraulic line failure in August 2000. As a result, a similar line failed on February 22, 2003, causing the operators to scram the reactor and trip the main turbine. The inspectors determined that the finding did not represent a noncompliance because it occurred on non-safety-related secondary plant equipment.
The inspectors determined that this problem identification and resolution finding is more than minor because the misdiagnosis of the August 31, 2000 failure contributed to the February 22, 2003, failure. The finding affected the initiating events cornerstone and was considered to have very low safety significance because it did not contribute to the likelihood of a LOCA, nor the likelihood of both a reactor scram and mitigating equipment or functions being unavailable, and because there was no increased likelihood of a fire or internal/external flood.
Inspection Report# : 2003006(pdf)
Mitigating Systems Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control conditions of engineered safety features electrical switchgear.
The inspectors identified two examples of a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to maintain the original design configuration of engineered safety feature switchgear. The inspectors found all of the heat dissipation louvers on top of the load centers and the relay control cabinets for both Divisions I and II auxiliary building 480 Vac engineered safety features switchgear covered with tape. Previously, the licensee had identified cardboard covering the ventilation louvers on breaker cubicles in the Division I engineered safety features 4160 Vac switchgear in the control building.
The failure to maintain design control over Switchgear EJS-SWGR2A and -2B and ENS-SWGR1A was a performance deficiency. The violation was more than minor because it was associated with the mitigating systems cornerstone attribute for design control. It affects the mitigating system cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. This noncited violation was evaluated using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." During the Phase 1 analysis, the issue was determined to have very low safety significance because it did not: (1) represent a design or qualification deficiency, (2) represent an actual loss of safety function of a system or a single train of a system for greater than the Technical Specification allowed out-of-service time, (3) represent an actual loss of safety function of non-
 
3Q/2004 Inspection Findings - River Bend 1                                                                                              Page 2 of 5 Technical Specification trains of equipment per 10 CFR 50.65 for more than 24 hours, and (4) screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Since this violation of 10 CFR Part 50, Appendix B, Criterion III, was of very low safety significance and was entered in the licensee's corrective action program as CR-RBS-2004-0512, -1389, -1855, and -1856, it is being treated as a noncited violation consistent with the NRC Enforcement Policy, NUREG-1600.
The inspectors also determined that on at least two occasions the licensee had the opportunity but failed to identify the tape covering the louvers on top of auxiliary building 480 Vac engineered safety features Switchgear EJS-SWGR2A. Therefore, the inspectors consider this finding to have problem identification and resolution crosscutting aspects for failure to identify a condition adverse to quality. Also the inspectors determined that the design engineering evaluation of as-found conditions for Division I engineered safety features 4160 Vac ENS-SWGR1A for past reportability was actually an evaluation of Division I 480 Vac engineered safety features EJS-SWGR1A and therefore a human performance error.
Inspection Report# : 2004003(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable preconditioning of Technical Specification diesel generator surveillance testing.
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's performance of unacceptable preconditioning of Technical Specification emergency diesel generator surveillance testing. The inspectors found three unacceptable preconditioning activities the licensee performed during the May and June 2004 emergency diesel generator monthly surveillance tests. The inspectors determined that this finding has problem identification and resolution aspects because the licensee identified some of these activities as unacceptable preconditioning in their evaluation of NRC Information Notice 97-16, "Preconditioning of Plant Structures, Systems, and Components Before ASME Code Inservice Testing or Technical Specification Surveillance Testing," dated June 9, 1997, yet failed to take actions to correct the test procedures.
The inspectors determined the unacceptable preconditioning of emergency diesel generator surveillance testing was a performance deficiency.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute for procedure quality. The finding affected the cornerstone objective to maintain availability and reliability of systems that respond to events to prevent undesirable consequences.
The inspectors reviewed the finding using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, was not an actual loss of safety function for a system or train, and was not risk significant due to a seismic, fire, flooding, or sever weather initiating event. The inspectors determined that unacceptable preconditioning of Technical Specification diesel generator surveillance testing was a violation of 10 CFR Part 50, Appendix B, Criterion V.
Because the violation was of very low safety significance and was entered into the licensee's corrective action program as CR-RBS-2004-1839 and -1858, it is being treated as an noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG 1600.
The inspectors identified crosscutting aspects related to problem identification and resolution. In their evaluation of NRC Information Notice 97-15, the licensee identified and evaluated some activities that precondition the emergency diesel generators during their prestart checks for surveillance testing, but failed to take appropriate actions to correct the procedures.
Inspection Report# : 2004003(pdf)
Significance:        Apr 06, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to adequately address roof leaks in the auxiliary building resulted in electrical grounds on safety related switchgear The licensee failed to adequately address leaks in the roof of the auxiliary building following several instances when roof leaks were identified and documented in the licensee's corrective action program. On February 5, 2004, rainwater inleakage through the auxiliary building roof resulted in an electrical ground on the control circuits of auxiliary building 480 Vac engineered safety features Switchgear EJS-SWG2A. The finding was of very low safety significance because, although it degraded one train of safety-related equipment, and could have degraded it again, it did not: increase the likelihood of a primary or secondary system loss of coolant accident initiator, contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, or increase the likelihood of a fire or internal/external flood.
The inspectors determined that the failure to correct the leaks in the auxiliary building was a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." Because this problem identification and resolution finding was of very low safety significance and was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-01083, it is being treated as an noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-16000.
Inspection Report# : 2004002(pdf)
Significance:        Feb 13, 2004 Identified By: Licensee Item Type: FIN Finding Untimely Corrective Actions for Degraded Fire Protection Feature
 
3Q/2004 Inspection Findings - River Bend 1                                                                                            Page 3 of 5 The licensee relied on compensatory measures for seven years instead of correcting a fire protection coating deficiency in three areas important to safe shutdown. In 1997, the licensee identified that the fire protective coatings on most structural steel beams in safety-related buildings did not meet the required thickness for a 3-hour fire rating. The deficient condition typically existed over one-fourth of each beam. While the majority of the deficiencies were repaired by building up the thickness, three fire areas remain degraded and had been subject to hourly fire watches since 1997. The team concluded that the planned corrective actions to restore the fire protection feature to its required condition for the remaining degraded areas were not timely.
This finding was greater than minor because it was similar to example 2.e in Appendix E of Manual Chapter 0609 and the finding is associated with degradation of a fire protection feature. This finding screened as having very low safety significance because the compensatory fire watches were in place as required and the remaining defense in depth elements remained unaffected.
Inspection Report# : 2004007(pdf)
Significance:        Feb 13, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Ventilation and Smoke Control Associated With a Fire The inspectors identified a non-cited violation of License Condition 2.C(10) and by reference the fire protection program and Appendix R to 10 CFR 50, Section III.K.12.h. The non-cited violation was identified related to fire response procedures and pre-fire strategies that did not contain adequate procedure steps for controlling the ventilation system alignment in order to both remove smoke and assure adequate cooling to remaining safe shutdown equipment. The team identified that the licensee did not account for fire dampers with heat-activated fusible links throughout the system, which could reasonably be expected to close when hot smoke was passed through the dampers. The licensee made a prompt change to FPP-0010, "Fire Fighting Procedure," to make operators aware of the condition as a compensatory measure. This issue was entered into the licensee's corrective action program under Condition Report 2004-000276.
This finding was greater than minor because it affected the Mitigating Systems Cornerstone objective of equipment reliability, in that loss of cooling or exposure to smoke and hot gases could cause failure of safe shutdown equipment that was supposed to remain unaffected by a particular fire. This finding screened as having very low safety significance because it affects a fire protection feature that was not a defense in depth element.
Inspection Report# : 2004007(pdf)
Barrier Integrity Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a functional failure.
The NRC inspectors identified a noncited violation of 10 CFR 50.65(a)(2). On May 15, 2003, the licensee failed to set goals and monitor the performance of the secondary containment system as required by 10 CFR 50.65(a)(1). As required by 10 CFR 50.65(a)(2), the licensee must demonstrate effective control of a structure's condition through appropriate preventive maintenance to not require paragraph (a)(1) monitoring.
The licensee had no justification for not requiring (a)(1) monitoring, after they failed to demonstrate effective control of the performance of the secondary containment system through appropriate preventive maintenance. The inspectors considered this violation to be noncited consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee entered this noncited violation into its corrective action program as Condition Report CR-RBS-2004-01706.
The inspectors determined this violation was more than minor because the failure to identify functional failures resulted in the system not being evaluated for 10 CFR 50.65(a)(1) status and had a credible impact on safety. The licensee performed engineering evaluations which concluded that, had a design basis accident occurred while the condition existed, the main control room, exclusion area boundary, and low population zone doses would have remained within the limits of 10CFR50.67. The inspectors determined the safety significance of this violation to be very low by the Reactor Safety Significance Determination Process. The inspectors answered the Phase 1 question regarding containment as yes because the inspectors determined that this finding represented a degradation of the radiological barrier only; therefore, in accordance with Manual Chapter 0609, Appendix A, Attachment 1, this finding is of very low safety significance.
Inspection Report# : 2004003(pdf)
Significance:        Nov 07, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Failed Open Secondary Containment Doors as Condition Adverse to Quality The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for failure to identify conditions that would have caused unexpected entry into Technical Specification Action Statements and had the potential to cause secondary containment to be inoperable.
 
3Q/2004 Inspection Findings - River Bend 1                                                                                            Page 4 of 5 The issue was more than minor because it affects the reactor safety/barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events. The results of the phase one evaluation of the significance determination process was that the issue was of very low safety significance because the finding only represents a degradation of the radiological barrier function provided by the auxiliary building and the duration of each of the 9 incidents was less than 10 minutes.
Inspection Report# : 2003007(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Mar 25, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify the correct configuration and adequacy of permanent shielding The inspectors identified a non-cited violation of Technical Specification 5.4.1.a because the licensee failed to follow procedural requirements to verify the correct configuration and adequacy of permanent shielding. On March 25, 2004, the inspectors identified that permanent shielding on a low-pressure core spray flush line, in the crescent area of the 70-foot elevation of the Auxiliary Building, was not in the correct configuration and not adequate for the intended application.
The failure to verify the correct configuration of permanent shielding and ensure that it was adequate for the intended application was 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. When processed through the Occupational Radiation Safety Significance Determination Process the finding was determined to be of very low safety significance because the finding was not associated with as low as is reasonably achievable issues, there was no overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. The finding was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-00924.
Inspection Report# : 2004002(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Nov 07, 2003 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Program Assessment The team concluded that the licensee was effective at identifying problems and putting them into the corrective action program. The licensee's effectiveness at problem identification was evidenced by the relatively few deficiencies identified by external organizations (including the NRC) that had not been previously identified by the licensee, during the review period. However, the team identified a repetitive failure on the part of the licensee to properly identify the inability of secondary containment doors to close and potential failures of secondary containment.
The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementing corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. On the basis of interviews conducted during this inspection, workers at the site felt free to input safety findings into the problem identification and resolution program.
 
3Q/2004 Inspection Findings - River Bend 1 Page 5 of 5 Inspection Report# : 2003007(pdf)
Last modified : December 29, 2004
 
4Q/2004 Inspection Findings - River Bend 1                                                                                              Page 1 of 7 River Bend 1 4Q/2004 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise a tagging boundary to support an emergent troubleshooting task resulted in a loss of protected division of offsite power and shutdown cooling The inspectors identified a green noncited violation of Technical Specification 5.4.1.a for failure to make a proper change to the tagging boundary around balance of plant Transformer RTX-XSR1F during Refueling Outage 12. This performance deficiency resulted in a trip signal, generated during troubleshooting the transformer sudden overpressure protection circuit, which caused the trip of switchyard Breakers OCB-20670 and OCB-20665. This resulted in the loss of offsite power to Division II engineered safety features Transformer RTX-XSR1D, causing a loss of shutdown cooling, a loss of alternate decay heat removal, containment isolations, and an automatic start of the Division II emergency diesel generator.
The inspectors determined that this human performance error was more than minor because it was associated with the initiating event cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the finding using IMC 0609, Appendix G, "Shutdown Operations Significance Determination Process,"
and determined that the loss of offsite power to Division II engineered safety features switchgear was of very low safety significance because there was no increased likelihood of a loss of reactor coolant system inventory, there was no loss of reactor water level instrumentation, there was no degradation of the licensee's ability to terminate a leak path or add water to the reactor when needed, nor was there any degradation of the licensee's ability to recover decay heat removal once it was lost. Because this human performance error was of very low safety significance (Green) and was documented in the licensee's corrective action program as CR-RBS-2003-03456, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Human performance error causes a loss of offsite power to Division I ESF Switchgear and start of the Division I emergency diesel generator during Refueling Outage 12 The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a. that was of very low safety significance (Green). As a result, during preparation for Division I integrated emergency core cooling systems testing, a technician inadvertently made contact with the wrong terminal on an undervoltage relay which tripped the preferred offsite power feeder breaker for the Division I safety-related 4160 Vac switchgear and started the Division I emergency diesel generator.
The inspectors determined that the inadvertent contact of the wrong terminal on Division I was a performance deficiency and a human performance error. Also, ineffective and incomplete corrective actions for similar errors contributed to the performance deficiency. The finding was more than minor because it was associated with the initiating events cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions, namely a partial loss of offsite power. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significant Determination Process," Attachment 1, Checklist 7, "BWR Refueling Operations with RCS Level greater than 23 feet." The finding was of very low safety significance (Green) because it did not cause a loss of shutdown cooling and did not compromise the ac power guidelines that: (1) one qualified circuit of offsite power remain operable; (2) at least one emergency diesel generator remain operable; and (3) necessary portions of the ac electrical power distribution systems remain operable.
The inspectors determined that this human performance error with problem identification and resolution aspects was the result of a violation of Technical Specification 5.4.1.a. which states, in part, that procedures shall be implemented and maintained as recommended in NUREG 1.33, Revision 2, Appendix A. Section 9.e. refers to general procedures for the control of maintenance activities. The licensee failed to evaluate the applicability of error reduction techniques, such as "taping of adjacent leads/contact points," for the installation of jumpers during Division I integrated emergency core cooling system testing, Procedure STP-309-0603, in accordance with Procedure ADM-0023, "Conduct of Maintenance," Revision 17A, Section 8.5. In addition, the licensee failed to install banana jacks on terminals on the back of the undervoltage relay in the Division I safety-related 4160 Vac switchgear, which were jumpered during the performance of Procedure STP-309-0603, in accordance with Procedure EDS-EE-001, "Banana Jack Standard," Revision 3. Because the finding was of very low safety significance and was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-3518, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
 
4Q/2004 Inspection Findings - River Bend 1                                                                                              Page 2 of 7 Significance:        Dec 31, 2004 Identified By: NRC Item Type: FIN Finding Automatic reactor scram during main turbine control valve testing due to control system malfunction The inspectors identified a finding based on the licensee's failure to adequately identify the root cause of the April 21, 2001, turbine trip and reactor scram so as to prevent recurrence. This failure resulted in a subsequent turbine trip and reactor scram on September 22, 2003.
The inspectors determined that the failure by the licensee to adequately identify the root cause of the April 21, 2001, event and to take effective corrective actions to prevent electrostatic arcing from affecting the primary and backup speed probes, was a performance deficiency. The inspectors determined that this performance deficiency led directly to the recurrence of the event on September 22, 2003. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affect mitigating equipment, and did not increase the likelihood of a fire or flood. This finding had problem identification and resolution crosscutting aspects regarding ineffective root cause determinations (evaluation). It was entered into the licensee's corrective action program as Condition Report CR-RBS-2003-3203.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding Failure to identify and properly evaluate deficient conditions related to switchyard breakers The inspectors identified a self-revealing finding of very low safety significance concerning the licensee's failure to identify a deficient condition due to preconditioned speed testing of station switchyard breakers and properly evaluate three similar failures of station switchyard breakers. As a result, three switchyard breakers opened slowly on August 15, 2004, and a transmission line ground fault that should have been isolated from the station switchyard remained connected to the main transformer long enough to cause a main generator lockout and reactor scram. Additionally, because slow breaker opening deenergized the north 230 kV bus, isolation of a coincident transmission line fault resulted in a loss of power to half of the balance of plant loads and the Division II engineered safety features switchboard.
This problem identification and resolution finding was more than minor because it was associated with the initiating events cornerstone objective to limit those events that upset plant stability and challenge a critical safety function during power operations. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because the finding contributed to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, the finding required a Phase 2 analysis. The inspectors referred the results of the Phase 2 analysis to the regional senior reactor analyst for final determination of risk.
The senior reactor analyst performed a Phase 3 analysis of the event. The factors that contributed to the result of that analysis included: (1) the dominant sequence was a transient with a loss of power to a vital bus; (2) the consequences of the finding were bounded by a complete loss of offsite power; (3) the history of single slow switchyard breaker operation; (4) the design and layout of the station switchyard; and (4) the possibility of recovery from either a partial or complete loss of offsite power given the conditions that led to the events of August 15, 2004. The result was that the finding was of very low safety significance.
Inspection Report# : 2004005(pdf)
Significance:        Oct 08, 2004 Identified By: Self Disclosing Item Type: FIN Finding Failure to maintain circulating water cooling tower drift eliminators and to take timely corrective actions to address insulator arcing The inspectors documented a self-revealing finding for failure to adequately maintain the circulating water cooling tower drift eliminators which resulted in salt contamination of the insulators in the on-site transformer yard, and failure to take corrective actions when pre-established trigger points were reached regarding insulator arcing (corona). The resulting contamination and failure to clean the insulators caused ground faults on Reserve Station Service Line1 and main transformers, which resulted in the loss of the Division I off-site power and a reactor scram on October 1, 2004. This finding had crosscutting aspects related to problem identification and resolution in that corrective actions were not implemented in a timely manner to prevent a significant plant transient.
This finding is more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. A completed Phase 3 evaluation resulted in an incremental conditional core damage probability of 1.2E-
: 7. Therefore, the significance of the finding was determined to be of very low safety significance.
Inspection Report# : 2004012(pdf)
 
4Q/2004 Inspection Findings - River Bend 1                                                                                              Page 3 of 7 Mitigating Systems Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Wide range reactor water level indication did not respond, as expected by operators, following an unplanned reactor scram A self-revealing, noncited violation of 10 CFR 55.46(c) was identified regarding differences between the simulator's and the plant's wide-range reactor water level digital indications during an unplanned reactor scram. This unexpected level indication resulted in indecision on the part of the operators during postscram recovery actions on December 10, 2004.
This finding is more than minor since deficiencies in the operator training program could become a more significant safety concern if left uncorrected. Based on the results of the significance determination process using Inspection Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process," this finding was determined to have very low safety significance, since it did not involve an exam or operating test but did involve a simulator fidelity issue which impacted operator actions during the response to an actual transient in the plant.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Rainwater leaked from auxiliary building roof onto Division I auxiliary building 480 Vac ESF switchgear, causing loss of a safety-related auxiliary building area unit The inspectors identified a self-revealing noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, for the licensee's failure to take timely and effective corrective action to prevent recurrence of rainwater leakage from the auxiliary building roof onto auxiliary building 480 Vac safety-related Switchgear EJS-SWGR2A, causing a loss of auxiliary building area unit Cooler HVR-UC11A. Investigation into the source of water determined that rainwater was accumulating inside the auxiliary building fresh air intake structure on the roof and leaking through seals along the air inlet ductwork onto Switchgear EJS-SWGR2A. The inspectors determined that this was a repeat of a February 5, 2004, leak documented in River Bend Station Condition Report 2004-0346 and a problem identification and resolution Noncited Violation 05000458/2004002-02. This finding had crosscutting aspects related to ineffective corrective actions.
The inspectors determined that the licensee's failure to take timely and effective corrective action to stop rainwater leaks from the auxiliary building roof onto Switchgear EJS-SWG2A was a performance deficiency that caused the loss of Cooler HVR-UC11A. The finding was more than minor because, if left uncorrected, rainwater leaks from the auxiliary building roof could lead to the loss of other Division I safety-related equipment and motor control centers powered by Switchgear EJS-SWG2A. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance because the short-term loss of unit Cooler HVR-UC11A did not cause an actual loss of safety function of any train of Technical Specification risk significant equipment and was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the failure to take timely and effective actions to prevent rainwater from leaking onto Switchgear EJS-SWGR2A was a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." Because this finding was of very low safety significance and was entered into the licensee's corrective action program as CR-RBS-2004-4218, this violation is being treated as a noncited violation, consistent with Section IV.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Oct 08, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to implement a required procedure for loss of main condenser vacuum/trip of circulating water pumps The inspectors identified a non-cited violation of Technical Specifications 5.4.1.a for the failure of the licensee to implement the Abnormal Operating Procedure AOP-0005, "Loss of Main Condenser Vacuum/Trip of Circulating Water Pump," following the loss of two of three operating circulating water pumps. Failure to implement this procedure contributed to the loss of condenser vacuum. This finding had cross-cutting aspects of human performance in that the operators did not implement the abnormal operating procedure as required. Additionally, this finding had cross-cutting aspects regarding problem identification and resolution in that a similar event had occurred over a month earlier, and no actions were taken to incorporate that operating experience into the operating procedures or process it through the corrective action program.
This finding is greater than minor because it is associated with human performance attribute of the mitigating system cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding actually led to the loss of main condenser vacuum and forced the operators to perform a reactor cool down through safety relief valves, reactor core isolation cooling and the suppression pool. This finding is of very low safety significance because it would only affect the plant during this particular situation of partial loss of offsite power and that all mitigating capability was maintained.
Inspection Report# : 2004012(pdf)
 
4Q/2004 Inspection Findings - River Bend 1                                                                                              Page 4 of 7 Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control conditions of engineered safety features electrical switchgear The inspectors identified two examples of a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to maintain the original design configuration of engineered safety feature switchgear. The inspectors found all of the heat dissipation louvers on top of the load centers and the relay control cabinets for both Divisions I and II auxiliary building 480 Vac engineered safety features switchgear covered with tape. Previously, the licensee had identified cardboard covering the ventilation louvers on breaker cubicles in the Division I engineered safety features 4160 Vac switchgear in the control building.
The failure to maintain design control over Switchgear EJS-SWGR2A and -2B and ENS-SWGR1A was a performance deficiency. The violation was more than minor because it was associated with the mitigating systems cornerstone attribute for design control. It affects the mitigating system cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. This noncited violation was evaluated using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." During the Phase 1 analysis, the issue was determined to have very low safety significance because it did not: (1) represent a design or qualification deficiency, (2) represent an actual loss of safety function of a system or a single train of a system for greater than the Technical Specification allowed out-of-service time, (3) represent an actual loss of safety function of non-Technical Specification trains of equipment per 10 CFR 50.65 for more than 24 hours, and (4) screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Since this violation of 10 CFR Part 50, Appendix B, Criterion III, was of very low safety significance and was entered in the licensee's corrective action program as CR-RBS-2004-0512, -1389, -1855, and -1856, it is being treated as a noncited violation consistent with the NRC Enforcement Policy, NUREG-1600.
The inspectors also determined that on at least two occasions the licensee had the opportunity but failed to identify the tape covering the louvers on top of auxiliary building 480 Vac engineered safety features Switchgear EJS-SWGR2A. Therefore, the inspectors consider this finding to have problem identification and resolution aspects for failure to identify a condition adverse to quality. Also the inspectors determined that the design engineering evaluation of as-found conditions for Division I engineered safety features 4160 Vac ENS-SWGR1A for past reportability was actually an evaluation of Division I 480 Vac engineered safety features EJS-SWGR1A and therefore a human performance error.
Inspection Report# : 2004003(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable preconditioning of Technical Specification diesel generator surveillance testing The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's performance of unacceptable preconditioning of Technical Specification emergency diesel generator surveillance testing. The inspectors found three unacceptable preconditioning activities the licensee performed during the May and June 2004 emergency diesel generator monthly surveillance tests. The inspectors determined that this finding has problem identification and resolution aspects because the licensee identified some of these activities as unacceptable preconditioning in their evaluation of NRC Information Notice 97-16, "Preconditioning of Plant Structures, Systems, and Components Before ASME Code Inservice Testing or Technical Specification Surveillance Testing," dated June 9, 1997, yet failed to take actions to correct the test procedures.
The inspectors determined the unacceptable preconditioning of emergency diesel generator surveillance testing was a performance deficiency.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute for procedure quality. The finding affected the cornerstone objective to maintain availability and reliability of systems that respond to events to prevent undesirable consequences.
The inspectors reviewed the finding using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, was not an actual loss of safety function for a system or train, and was not risk significant due to a seismic, fire, flooding, or severe weather initiating event. The inspectors determined that unacceptable preconditioning of Technical Specification diesel generator surveillance testing was a violation of 10 CFR Part 50, Appendix B, Criterion V. Because the violation was of very low safety significance and was entered into the licensee's corrective action program as CR-RBS-2004-1839 and -1858, it is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG 1600.
The inspectors identified aspects related to problem identification and resolution. In their evaluation of NRC Information Notice 97-15, the licensee identified and evaluated some activities that precondition the emergency diesel generators during their prestart checks for surveillance testing, but failed to take appropriate actions to correct the procedures.
Inspection Report# : 2004003(pdf)
Significance:        Mar 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to adequately address roof leaks in the auxiliary building resulted in electrical grounds on safety-related switchgear
 
4Q/2004 Inspection Findings - River Bend 1                                                                                              Page 5 of 7 The licensee failed to adequately address leaks in the roof of the auxiliary building following several instances when roof leaks were identified and documented in the licensee's corrective action program. On February 5, 2004, rainwater inleakage through the auxiliary building roof resulted in an electrical ground on the control circuits of auxiliary building 480 Vac engineered safety features Switchgear EJS-SWG2A. The finding was of very low safety significance because, although it degraded one train of safety-related equipment, and could have degraded it again, it did not: increase the likelihood of a primary or secondary system loss of coolant accident initiator, contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, or increase the likelihood of a fire or internal/external flood.
The inspectors determined that the failure to correct the leaks in the auxiliary building was a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." Because this problem identification and resolution finding was of very low safety significance and was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-01083, it is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004002(pdf)
Significance:        Feb 13, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedures for ventilation and smoke control associated with a fire The inspectors identified a noncited violation of License Condition 2.C(10) and by reference the fire protection program and Appendix R to 10 CFR 50, Section III.K.12.h. The noncited violation was identified related to fire response procedures and pre-fire strategies that did not contain adequate procedure steps for controlling the ventilation system alignment in order to both remove smoke and assure adequate cooling to remaining safe shutdown equipment. The team identified that the licensee did not account for fire dampers with heat-activated fusible links throughout the system, which could reasonably be expected to close when hot smoke was passed through the dampers. The licensee made a prompt change to FPP-0010, "Fire Fighting Procedure," to make operators aware of the condition as a compensatory measure. This issue was entered into the licensee's corrective action program under Condition Report 2004-000276.
This finding was greater than minor because it affected the Mitigating Systems Cornerstone objective of equipment reliability, in that loss of cooling or exposure to smoke and hot gases could cause failure of safe shutdown equipment that was supposed to remain unaffected by a particular fire. This finding screened as having very low safety significance because it affects a fire protection feature that was not a defense in depth element.
Inspection Report# : 2004007(pdf)
Significance:        Feb 13, 2004 Identified By: Licensee Item Type: FIN Finding Untimely corrective actions for degraded fire protection feature The licensee relied on compensatory measures for seven years instead of correcting a fire protection coating deficiency in three areas important to safe shutdown. In 1997, the licensee identified that the fire protective coatings on most structural steel beams in safety-related buildings did not meet the required thickness for a 3-hour fire rating. The deficient condition typically existed over one-fourth of each beam. While the majority of the deficiencies were repaired by building up the thickness, three fire areas remain degraded and had been subject to hourly fire watches since 1997. The team concluded that the planned corrective actions to restore the fire protection feature to its required condition for the remaining degraded areas were not timely.
This finding was greater than minor because it was similar to example 2.e in Appendix E of Manual Chapter 0609 and the finding is associated with degradation of a fire protection feature. This finding screened as having very low safety significance because the compensatory fire watches were in place as required and the remaining defense in depth elements remained unaffected.
Inspection Report# : 2004007(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to control special processes such as welding in accordance with qualified welding procedures The inspector identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion IX, for failure to control special processes, such as welding, in accordance with qualified welding procedures as required. The finding was a human performance error for the failure to follow procedure. Criterion IX, Appendix B, of 10 CFR Part 50, "Control of Special Processes," requires in part that measures shall be established to assure that special processes, including welding, heat treating, and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
Contrary to the above, welding personnel failed to verify interpass temperature during welding activities on feedwater inlet check Valve B21-
 
4Q/2004 Inspection Findings - River Bend 1                                                                                            Page 6 of 7 AOVF032, an ASME Class1 valve, in accordance with qualified welding procedures.
This finding was determined to be more than minor, through Inspection Manual Chapter 0612, Appendix B, in that it affected the barrier integrity cornerstone attribute of human performance, could have represented a more significant issue if left uncorrected, and there was a reasonable likelihood that the valve would have been returned to service if the inspector had not intervened. Based on the results of a significance determination process Phase 1 analysis, this finding had very low safety significance because it did not result in the loss of a barrier integrity function and has been entered into the licensee's corrective action program as Condition Report CR-RBS-2004-03395. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a functional failure The NRC inspectors identified a noncited violation of 10 CFR 50.65(a)(2). On May 15, 2003, the licensee failed to set goals and monitor the performance of the secondary containment system as required by 10 CFR 50.65(a)(1). As required by 10 CFR 50.65(a)(2), the licensee must demonstrate effective control of a structure's condition through appropriate preventive maintenance to not require paragraph (a)(1) monitoring.
The licensee had no justification for not requiring (a)(1) monitoring, after they failed to demonstrate effective control of the performance of the secondary containment system through appropriate preventive maintenance. The inspectors considered this violation to be noncited consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee entered this noncited violation into its corrective action program as Condition Report CR-RBS-2004-01706.
The inspectors determined this violation was more than minor because the failure to identify functional failures resulted in the system not being evaluated for 10 CFR 50.65(a)(1) status and had a credible impact on safety. The licensee performed engineering evaluations which concluded that, had a design basis accident occurred while the condition existed, the main control room, exclusion area boundary, and low population zone doses would have remained within the limits of 10CFR50.67. The inspectors determined the safety significance of this violation to be very low by the Reactor Safety Significance Determination Process. The inspectors answered the Phase 1 question regarding containment as yes because the inspectors determined that this finding represented a degradation of the radiological barrier only; therefore, in accordance with Manual Chapter 0609, Appendix A, Attachment 1, this finding is of very low safety significance.
Inspection Report# : 2004003(pdf)
Significance:        Mar 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Reactor operated in excess of licensed maximum power level due to incorrect feedwater flow calculations The licensee operated the reactor plant at power levels above the licensed maximum power level from February 1996 to May 2003 due to an error in feedwater flow rate used to calculate reactor core thermal power. It was found that the feedwater flow rate data was inaccurate by as much as 2.69 percent rated system flow and actual thermal power was as much as 2.7 percent higher than the calculated thermal power. The inspectors determined that this finding was a problem identification and resolution finding because the licensee missed several opportunities to identify and correct this overpower condition.
The finding was more than minor because if left uncorrected and a design basis accident occurred the resulting fuel damage could exceed analyzed values. The inspectors determined that the finding affected the reactor fuel cladding barrier, but was of very low safety significance because the reactor coolant system barrier was not effected. This self-revealing finding was a violation of operating license Condition 2.C.(1),
"Maximum Power Level." Because the violation was of very low safety significance and was entered in the licensee's corrective action program as Condition Report CR-RBS-2003-02082, it is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004002(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation
 
4Q/2004 Inspection Findings - River Bend 1                                                                                            Page 7 of 7 Failure to control a high radiation area in accordance with Technical Specification 5.7.3 The inspector reviewed a self-revealing noncited violation of Technical Specification 5.7.3 because the licensee failed to control a high radiation area with dose rates greater than 1,000 millirems per hour. On October 31, 2004, during maintenance activities on valves located on the 82-foot level of the drywell, three workers' electronic alarming dosimeters unexpectedly alarmed when they were exposed to unanticipated radiation levels of approximately 1,700 millirems per hour. Subsequent radiation surveys at the source of radiation around Valve RCS-V-3009 identified 6,000 millirems per hour on contact and 2,000 millirems per hour at 30 centimeters. The area was not barricaded, conspicuously posted, and did not have a flashing light activated as a warning device. The licensee determined that the three workers received 84, 85, and 95 millirems, respectively. This finding was entered into the licensee's corrective action program.
This finding is more than minor because it is associated with the Occupational Radiation Safety attribute of exposure control and affected the cornerstone objective, in that not controlling locked high radiation areas could increase personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance (Green) 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.
Inspection Report# : 2004005(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify the correct configuration and adequacy of permanent shielding The inspectors identified a noncited violation of Technical Specification 5.4.1.a because the licensee failed to follow procedural requirements to verify the correct configuration and adequacy of permanent shielding. On March 25, 2004, the inspectors identified that permanent shielding on a low-pressure core spray flush line, in the crescent area of the 70-foot elevation of the auxiliary building, was not in the correct configuration and not adequate for the intended application.
The failure to verify the correct configuration of permanent shielding and ensure that it was adequate for the intended application was 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 effected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. When processed through the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because the finding was not associated with as low as is reasonably achievable issues, there was no overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. The finding was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-00924.
Inspection Report# : 2004002(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : March 09, 2005
 
1Q/2005 Inspection Findings - River Bend 1                                                                                              Page 1 of 6 River Bend 1 1Q/2005 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise a tagging boundary to support an emergent troubleshooting task resulted in a loss of protected division of offsite power and shutdown cooling The inspectors identified a green noncited violation of Technical Specification 5.4.1.a for failure to make a proper change to the tagging boundary around balance of plant Transformer RTX-XSR1F during Refueling Outage 12. This performance deficiency resulted in a trip signal, generated during troubleshooting the transformer sudden overpressure protection circuit, which caused the trip of switchyard Breakers OCB-20670 and OCB-20665. This resulted in the loss of offsite power to Division II engineered safety features Transformer RTX-XSR1D, causing a loss of shutdown cooling, a loss of alternate decay heat removal, containment isolations, and an automatic start of the Division II emergency diesel generator.
The inspectors determined that this human performance error was more than minor because it was associated with the initiating event cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the finding using IMC 0609, Appendix G, "Shutdown Operations Significance Determination Process,"
and determined that the loss of offsite power to Division II engineered safety features switchgear was of very low safety significance because there was no increased likelihood of a loss of reactor coolant system inventory, there was no loss of reactor water level instrumentation, there was no degradation of the licensee's ability to terminate a leak path or add water to the reactor when needed, nor was there any degradation of the licensee's ability to recover decay heat removal once it was lost. Because this human performance error was of very low safety significance (Green) and was documented in the licensee's corrective action program as CR-RBS-2003-03456, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Human performance error causes a loss of offsite power to Division I ESF Switchgear and start of the Division I emergency diesel generator during Refueling Outage 12 The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a. that was of very low safety significance (Green). As a result, during preparation for Division I integrated emergency core cooling systems testing, a technician inadvertently made contact with the wrong terminal on an undervoltage relay which tripped the preferred offsite power feeder breaker for the Division I safety-related 4160 Vac switchgear and started the Division I emergency diesel generator.
The inspectors determined that the inadvertent contact of the wrong terminal on Division I was a performance deficiency and a human performance error. Also, ineffective and incomplete corrective actions for similar errors contributed to the performance deficiency. The finding was more than minor because it was associated with the initiating events cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions, namely a partial loss of offsite power. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significant Determination Process," Attachment 1, Checklist 7, "BWR Refueling Operations with RCS Level greater than 23 feet." The finding was of very low safety significance (Green) because it did not cause a loss of shutdown cooling and did not compromise the ac power guidelines that: (1) one qualified circuit of offsite power remain operable; (2) at least one emergency diesel generator remain operable; and (3) necessary portions of the ac electrical power distribution systems remain operable.
The inspectors determined that this human performance error with problem identification and resolution aspects was the result of a violation of Technical Specification 5.4.1.a. which states, in part, that procedures shall be implemented and maintained as recommended in NUREG 1.33, Revision 2, Appendix A. Section 9.e. refers to general procedures for the control of maintenance activities. The licensee failed to evaluate the applicability of error reduction techniques, such as "taping of adjacent leads/contact points," for the installation of jumpers during Division I integrated emergency core cooling system testing, Procedure STP-309-0603, in accordance with Procedure ADM-0023, "Conduct of Maintenance," Revision 17A, Section 8.5. In addition, the licensee failed to install banana jacks on terminals on the back of the undervoltage relay in the Division I safety-related 4160 Vac switchgear, which were jumpered during the performance of Procedure STP-309-0603, in accordance with Procedure EDS-EE-001, "Banana Jack Standard," Revision 3. Because the finding was of very low safety significance and was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-3518, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
 
1Q/2005 Inspection Findings - River Bend 1                                                                                              Page 2 of 6 Significance:        Dec 31, 2004 Identified By: NRC Item Type: FIN Finding Automatic reactor scram during main turbine control valve testing due to control system malfunction The inspectors identified a finding based on the licensee's failure to adequately identify the root cause of the April 21, 2001, turbine trip and reactor scram so as to prevent recurrence. This failure resulted in a subsequent turbine trip and reactor scram on September 22, 2003.
The inspectors determined that the failure by the licensee to adequately identify the root cause of the April 21, 2001, event and to take effective corrective actions to prevent electrostatic arcing from affecting the primary and backup speed probes, was a performance deficiency. The inspectors determined that this performance deficiency led directly to the recurrence of the event on September 22, 2003. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affect mitigating equipment, and did not increase the likelihood of a fire or flood. This finding had problem identification and resolution crosscutting aspects regarding ineffective root cause determinations (evaluation). It was entered into the licensee's corrective action program as Condition Report CR-RBS-2003-3203.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding Failure to identify and properly evaluate deficient conditions related to switchyard breakers The inspectors identified a self-revealing finding of very low safety significance concerning the licensee's failure to identify a deficient condition due to preconditioned speed testing of station switchyard breakers and properly evaluate three similar failures of station switchyard breakers. As a result, three switchyard breakers opened slowly on August 15, 2004, and a transmission line ground fault that should have been isolated from the station switchyard remained connected to the main transformer long enough to cause a main generator lockout and reactor scram. Additionally, because slow breaker opening deenergized the north 230 kV bus, isolation of a coincident transmission line fault resulted in a loss of power to half of the balance of plant loads and the Division II engineered safety features switchboard.
This problem identification and resolution finding was more than minor because it was associated with the initiating events cornerstone objective to limit those events that upset plant stability and challenge a critical safety function during power operations. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because the finding contributed to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, the finding required a Phase 2 analysis. The inspectors referred the results of the Phase 2 analysis to the regional senior reactor analyst for final determination of risk.
The senior reactor analyst performed a Phase 3 analysis of the event. The factors that contributed to the result of that analysis included: (1) the dominant sequence was a transient with a loss of power to a vital bus; (2) the consequences of the finding were bounded by a complete loss of offsite power; (3) the history of single slow switchyard breaker operation; (4) the design and layout of the station switchyard; and (4) the possibility of recovery from either a partial or complete loss of offsite power given the conditions that led to the events of August 15, 2004. The result was that the finding was of very low safety significance.
Inspection Report# : 2004005(pdf)
Significance:        Oct 08, 2004 Identified By: Self Disclosing Item Type: FIN Finding Failure to maintain circulating water cooling tower drift eliminators and to take timely corrective actions to address insulator arcing The inspectors documented a self-revealing finding for failure to adequately maintain the circulating water cooling tower drift eliminators which resulted in salt contamination of the insulators in the on-site transformer yard, and failure to take corrective actions when pre-established trigger points were reached regarding insulator arcing (corona). The resulting contamination and failure to clean the insulators caused ground faults on Reserve Station Service Line1 and main transformers, which resulted in the loss of the Division I off-site power and a reactor scram on October 1, 2004. This finding had crosscutting aspects related to problem identification and resolution in that corrective actions were not implemented in a timely manner to prevent a significant plant transient.
This finding is more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. A completed Phase 3 evaluation resulted in an incremental conditional core damage probability of 1.2E-
: 7. Therefore, the significance of the finding was determined to be of very low safety significance.
Inspection Report# : 2004012(pdf)
 
1Q/2005 Inspection Findings - River Bend 1                                                                                              Page 3 of 6 Mitigating Systems Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Wide range reactor water level indication did not respond, as expected by operators, following an unplanned reactor scram A self-revealing, noncited violation of 10 CFR 55.46(c) was identified regarding differences between the simulator's and the plant's wide-range reactor water level digital indications during an unplanned reactor scram. This unexpected level indication resulted in indecision on the part of the operators during postscram recovery actions on December 10, 2004.
This finding is more than minor since deficiencies in the operator training program could become a more significant safety concern if left uncorrected. Based on the results of the significance determination process using Inspection Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process," this finding was determined to have very low safety significance, since it did not involve an exam or operating test but did involve a simulator fidelity issue which impacted operator actions during the response to an actual transient in the plant.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Rainwater leaked from auxiliary building roof onto Division I auxiliary building 480 Vac ESF switchgear, causing loss of a safety-related auxiliary building area unit The inspectors identified a self-revealing noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, for the licensee's failure to take timely and effective corrective action to prevent recurrence of rainwater leakage from the auxiliary building roof onto auxiliary building 480 Vac safety-related Switchgear EJS-SWGR2A, causing a loss of auxiliary building area unit Cooler HVR-UC11A. Investigation into the source of water determined that rainwater was accumulating inside the auxiliary building fresh air intake structure on the roof and leaking through seals along the air inlet ductwork onto Switchgear EJS-SWGR2A. The inspectors determined that this was a repeat of a February 5, 2004, leak documented in River Bend Station Condition Report 2004-0346 and a problem identification and resolution Noncited Violation 05000458/2004002-02. This finding had crosscutting aspects related to ineffective corrective actions.
The inspectors determined that the licensee's failure to take timely and effective corrective action to stop rainwater leaks from the auxiliary building roof onto Switchgear EJS-SWG2A was a performance deficiency that caused the loss of Cooler HVR-UC11A. The finding was more than minor because, if left uncorrected, rainwater leaks from the auxiliary building roof could lead to the loss of other Division I safety-related equipment and motor control centers powered by Switchgear EJS-SWG2A. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance because the short-term loss of unit Cooler HVR-UC11A did not cause an actual loss of safety function of any train of Technical Specification risk significant equipment and was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the failure to take timely and effective actions to prevent rainwater from leaking onto Switchgear EJS-SWGR2A was a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." Because this finding was of very low safety significance and was entered into the licensee's corrective action program as CR-RBS-2004-4218, this violation is being treated as a noncited violation, consistent with Section IV.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Oct 08, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to implement a required procedure for loss of main condenser vacuum/trip of circulating water pumps The inspectors identified a non-cited violation of Technical Specifications 5.4.1.a for the failure of the licensee to implement the Abnormal Operating Procedure AOP-0005, "Loss of Main Condenser Vacuum/Trip of Circulating Water Pump," following the loss of two of three operating circulating water pumps. Failure to implement this procedure contributed to the loss of condenser vacuum. This finding had cross-cutting aspects of human performance in that the operators did not implement the abnormal operating procedure as required. Additionally, this finding had cross-cutting aspects regarding problem identification and resolution in that a similar event had occurred over a month earlier, and no actions were taken to incorporate that operating experience into the operating procedures or process it through the corrective action program.
This finding is greater than minor because it is associated with human performance attribute of the mitigating system cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding actually led to the loss of main condenser vacuum and forced the operators to perform a reactor cool down through safety relief valves, reactor core isolation cooling and the suppression pool. This finding is of very low safety significance because it would only affect the plant during this particular situation of partial loss of offsite power and that all mitigating capability was maintained.
Inspection Report# : 2004012(pdf)
 
1Q/2005 Inspection Findings - River Bend 1                                                                                              Page 4 of 6 Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control conditions of engineered safety features electrical switchgear The inspectors identified two examples of a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to maintain the original design configuration of engineered safety feature switchgear. The inspectors found all of the heat dissipation louvers on top of the load centers and the relay control cabinets for both Divisions I and II auxiliary building 480 Vac engineered safety features switchgear covered with tape. Previously, the licensee had identified cardboard covering the ventilation louvers on breaker cubicles in the Division I engineered safety features 4160 Vac switchgear in the control building.
The failure to maintain design control over Switchgear EJS-SWGR2A and -2B and ENS-SWGR1A was a performance deficiency. The violation was more than minor because it was associated with the mitigating systems cornerstone attribute for design control. It affects the mitigating system cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. This noncited violation was evaluated using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." During the Phase 1 analysis, the issue was determined to have very low safety significance because it did not: (1) represent a design or qualification deficiency, (2) represent an actual loss of safety function of a system or a single train of a system for greater than the Technical Specification allowed out-of-service time, (3) represent an actual loss of safety function of non-Technical Specification trains of equipment per 10 CFR 50.65 for more than 24 hours, and (4) screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event. Since this violation of 10 CFR Part 50, Appendix B, Criterion III, was of very low safety significance and was entered in the licensee's corrective action program as CR-RBS-2004-0512, -1389, -1855, and -1856, it is being treated as a noncited violation consistent with the NRC Enforcement Policy, NUREG-1600.
The inspectors also determined that on at least two occasions the licensee had the opportunity but failed to identify the tape covering the louvers on top of auxiliary building 480 Vac engineered safety features Switchgear EJS-SWGR2A. Therefore, the inspectors consider this finding to have problem identification and resolution aspects for failure to identify a condition adverse to quality. Also the inspectors determined that the design engineering evaluation of as-found conditions for Division I engineered safety features 4160 Vac ENS-SWGR1A for past reportability was actually an evaluation of Division I 480 Vac engineered safety features EJS-SWGR1A and therefore a human performance error.
Inspection Report# : 2004003(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable preconditioning of Technical Specification diesel generator surveillance testing The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's performance of unacceptable preconditioning of Technical Specification emergency diesel generator surveillance testing. The inspectors found three unacceptable preconditioning activities the licensee performed during the May and June 2004 emergency diesel generator monthly surveillance tests. The inspectors determined that this finding has problem identification and resolution aspects because the licensee identified some of these activities as unacceptable preconditioning in their evaluation of NRC Information Notice 97-16, "Preconditioning of Plant Structures, Systems, and Components Before ASME Code Inservice Testing or Technical Specification Surveillance Testing," dated June 9, 1997, yet failed to take actions to correct the test procedures.
The inspectors determined the unacceptable preconditioning of emergency diesel generator surveillance testing was a performance deficiency.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute for procedure quality. The finding affected the cornerstone objective to maintain availability and reliability of systems that respond to events to prevent undesirable consequences.
The inspectors reviewed the finding using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, was not an actual loss of safety function for a system or train, and was not risk significant due to a seismic, fire, flooding, or severe weather initiating event. The inspectors determined that unacceptable preconditioning of Technical Specification diesel generator surveillance testing was a violation of 10 CFR Part 50, Appendix B, Criterion V. Because the violation was of very low safety significance and was entered into the licensee's corrective action program as CR-RBS-2004-1839 and -1858, it is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG 1600.
The inspectors identified aspects related to problem identification and resolution. In their evaluation of NRC Information Notice 97-15, the licensee identified and evaluated some activities that precondition the emergency diesel generators during their prestart checks for surveillance testing, but failed to take appropriate actions to correct the procedures.
Inspection Report# : 2004003(pdf)
Barrier Integrity
 
1Q/2005 Inspection Findings - River Bend 1                                                                                            Page 5 of 6 Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to control special processes such as welding in accordance with qualified welding procedures The inspector identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion IX, for failure to control special processes, such as welding, in accordance with qualified welding procedures as required. The finding was a human performance error for the failure to follow procedure. Criterion IX, Appendix B, of 10 CFR Part 50, "Control of Special Processes," requires in part that measures shall be established to assure that special processes, including welding, heat treating, and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
Contrary to the above, welding personnel failed to verify interpass temperature during welding activities on feedwater inlet check Valve B21-AOVF032, an ASME Class1 valve, in accordance with qualified welding procedures.
This finding was determined to be more than minor, through Inspection Manual Chapter 0612, Appendix B, in that it affected the barrier integrity cornerstone attribute of human performance, could have represented a more significant issue if left uncorrected, and there was a reasonable likelihood that the valve would have been returned to service if the inspector had not intervened. Based on the results of a significance determination process Phase 1 analysis, this finding had very low safety significance because it did not result in the loss of a barrier integrity function and has been entered into the licensee's corrective action program as Condition Report CR-RBS-2004-03395. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify a functional failure The NRC inspectors identified a noncited violation of 10 CFR 50.65(a)(2). On May 15, 2003, the licensee failed to set goals and monitor the performance of the secondary containment system as required by 10 CFR 50.65(a)(1). As required by 10 CFR 50.65(a)(2), the licensee must demonstrate effective control of a structure's condition through appropriate preventive maintenance to not require paragraph (a)(1) monitoring.
The licensee had no justification for not requiring (a)(1) monitoring, after they failed to demonstrate effective control of the performance of the secondary containment system through appropriate preventive maintenance. The inspectors considered this violation to be noncited consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee entered this noncited violation into its corrective action program as Condition Report CR-RBS-2004-01706.
The inspectors determined this violation was more than minor because the failure to identify functional failures resulted in the system not being evaluated for 10 CFR 50.65(a)(1) status and had a credible impact on safety. The licensee performed engineering evaluations which concluded that, had a design basis accident occurred while the condition existed, the main control room, exclusion area boundary, and low population zone doses would have remained within the limits of 10CFR50.67. The inspectors determined the safety significance of this violation to be very low by the Reactor Safety Significance Determination Process. The inspectors answered the Phase 1 question regarding containment as yes because the inspectors determined that this finding represented a degradation of the radiological barrier only; therefore, in accordance with Manual Chapter 0609, Appendix A, Attachment 1, this finding is of very low safety significance.
Inspection Report# : 2004003(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to control a high radiation area in accordance with Technical Specification 5.7.3 The inspector reviewed a self-revealing noncited violation of Technical Specification 5.7.3 because the licensee failed to control a high radiation area with dose rates greater than 1,000 millirems per hour. On October 31, 2004, during maintenance activities on valves located on the 82-foot level of the drywell, three workers' electronic alarming dosimeters unexpectedly alarmed when they were exposed to unanticipated radiation levels of approximately 1,700 millirems per hour. Subsequent radiation surveys at the source of radiation around Valve RCS-V-3009 identified 6,000 millirems per hour on contact and 2,000 millirems per hour at 30 centimeters. The area was not barricaded, conspicuously posted, and did not have a flashing light activated as a warning device. The licensee determined that the three workers received 84, 85, and 95 millirems, respectively. This finding was entered into the licensee's corrective action program.
 
1Q/2005 Inspection Findings - River Bend 1                                                                                          Page 6 of 6 This finding is more than minor because it is associated with the Occupational Radiation Safety attribute of exposure control and affected the cornerstone objective, in that not controlling locked high radiation areas could increase personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance (Green) 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.
Inspection Report# : 2004005(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : June 17, 2005
 
2Q/2005 Inspection Findings - River Bend 1                                                                                              Page 1 of 5 River Bend 1 2Q/2005 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise a tagging boundary to support an emergent troubleshooting task resulted in a loss of protected division of offsite power and shutdown cooling The inspectors identified a green noncited violation of Technical Specification 5.4.1.a for failure to make a proper change to the tagging boundary around balance of plant Transformer RTX-XSR1F during Refueling Outage 12. This performance deficiency resulted in a trip signal, generated during troubleshooting the transformer sudden overpressure protection circuit, which caused the trip of switchyard Breakers OCB-20670 and OCB-20665. This resulted in the loss of offsite power to Division II engineered safety features Transformer RTX-XSR1D, causing a loss of shutdown cooling, a loss of alternate decay heat removal, containment isolations, and an automatic start of the Division II emergency diesel generator.
The inspectors determined that this human performance error was more than minor because it was associated with the initiating event cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the finding using IMC 0609, Appendix G, "Shutdown Operations Significance Determination Process,"
and determined that the loss of offsite power to Division II engineered safety features switchgear was of very low safety significance because there was no increased likelihood of a loss of reactor coolant system inventory, there was no loss of reactor water level instrumentation, there was no degradation of the licensee's ability to terminate a leak path or add water to the reactor when needed, nor was there any degradation of the licensee's ability to recover decay heat removal once it was lost. Because this human performance error was of very low safety significance (Green) and was documented in the licensee's corrective action program as CR-RBS-2003-03456, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Human performance error causes a loss of offsite power to Division I ESF Switchgear and start of the Division I emergency diesel generator during Refueling Outage 12 The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a. that was of very low safety significance (Green). As a result, during preparation for Division I integrated emergency core cooling systems testing, a technician inadvertently made contact with the wrong terminal on an undervoltage relay which tripped the preferred offsite power feeder breaker for the Division I safety-related 4160 Vac switchgear and started the Division I emergency diesel generator.
The inspectors determined that the inadvertent contact of the wrong terminal on Division I was a performance deficiency and a human performance error. Also, ineffective and incomplete corrective actions for similar errors contributed to the performance deficiency. The finding was more than minor because it was associated with the initiating events cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions, namely a partial loss of offsite power. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significant Determination Process," Attachment 1, Checklist 7, "BWR Refueling Operations with RCS Level greater than 23 feet." The finding was of very low safety significance (Green) because it did not cause a loss of shutdown cooling and did not compromise the ac power guidelines that: (1) one qualified circuit of offsite power remain operable; (2) at least one emergency diesel generator remain operable; and (3) necessary portions of the ac electrical power distribution systems remain operable.
The inspectors determined that this human performance error with problem identification and resolution aspects was the result of a violation of Technical Specification 5.4.1.a. which states, in part, that procedures shall be implemented and maintained as recommended in NUREG 1.33, Revision 2, Appendix A. Section 9.e. refers to general procedures for the control of maintenance activities. The licensee failed to evaluate the applicability of error reduction techniques, such as "taping of adjacent leads/contact points," for the installation of jumpers during Division I integrated emergency core cooling system testing, Procedure STP-309-0603, in accordance with Procedure ADM-0023, "Conduct of Maintenance," Revision 17A, Section 8.5. In addition, the licensee failed to install banana jacks on terminals on the back of the undervoltage relay in the Division I safety-related 4160 Vac switchgear, which were jumpered during the performance of Procedure STP-309-0603, in accordance with Procedure EDS-EE-001, "Banana Jack Standard," Revision 3. Because the finding was of very low safety significance and was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-3518, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
 
2Q/2005 Inspection Findings - River Bend 1                                                                                              Page 2 of 5 Significance:        Dec 31, 2004 Identified By: NRC Item Type: FIN Finding Automatic reactor scram during main turbine control valve testing due to control system malfunction The inspectors identified a finding based on the licensee's failure to adequately identify the root cause of the April 21, 2001, turbine trip and reactor scram so as to prevent recurrence. This failure resulted in a subsequent turbine trip and reactor scram on September 22, 2003.
The inspectors determined that the failure by the licensee to adequately identify the root cause of the April 21, 2001, event and to take effective corrective actions to prevent electrostatic arcing from affecting the primary and backup speed probes, was a performance deficiency. The inspectors determined that this performance deficiency led directly to the recurrence of the event on September 22, 2003. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affect mitigating equipment, and did not increase the likelihood of a fire or flood. This finding had problem identification and resolution crosscutting aspects regarding ineffective root cause determinations (evaluation). It was entered into the licensee's corrective action program as Condition Report CR-RBS-2003-3203.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: FIN Finding Failure to identify and properly evaluate deficient conditions related to switchyard breakers The inspectors identified a self-revealing finding of very low safety significance concerning the licensee's failure to identify a deficient condition due to preconditioned speed testing of station switchyard breakers and properly evaluate three similar failures of station switchyard breakers. As a result, three switchyard breakers opened slowly on August 15, 2004, and a transmission line ground fault that should have been isolated from the station switchyard remained connected to the main transformer long enough to cause a main generator lockout and reactor scram. Additionally, because slow breaker opening deenergized the north 230 kV bus, isolation of a coincident transmission line fault resulted in a loss of power to half of the balance of plant loads and the Division II engineered safety features switchboard.
This problem identification and resolution finding was more than minor because it was associated with the initiating events cornerstone objective to limit those events that upset plant stability and challenge a critical safety function during power operations. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because the finding contributed to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, the finding required a Phase 2 analysis. The inspectors referred the results of the Phase 2 analysis to the regional senior reactor analyst for final determination of risk.
The senior reactor analyst performed a Phase 3 analysis of the event. The factors that contributed to the result of that analysis included: (1) the dominant sequence was a transient with a loss of power to a vital bus; (2) the consequences of the finding were bounded by a complete loss of offsite power; (3) the history of single slow switchyard breaker operation; (4) the design and layout of the station switchyard; and (4) the possibility of recovery from either a partial or complete loss of offsite power given the conditions that led to the events of August 15, 2004. The result was that the finding was of very low safety significance.
Inspection Report# : 2004005(pdf)
Significance:        Oct 08, 2004 Identified By: Self Disclosing Item Type: FIN Finding Failure to maintain circulating water cooling tower drift eliminators and to take timely corrective actions to address insulator arcing The inspectors documented a self-revealing finding for failure to adequately maintain the circulating water cooling tower drift eliminators which resulted in salt contamination of the insulators in the on-site transformer yard, and failure to take corrective actions when pre-established trigger points were reached regarding insulator arcing (corona). The resulting contamination and failure to clean the insulators caused ground faults on Reserve Station Service Line1 and main transformers, which resulted in the loss of the Division I off-site power and a reactor scram on October 1, 2004. This finding had crosscutting aspects related to problem identification and resolution in that corrective actions were not implemented in a timely manner to prevent a significant plant transient.
This finding is more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. A completed Phase 3 evaluation resulted in an incremental conditional core damage probability of 1.2E-
: 7. Therefore, the significance of the finding was determined to be of very low safety significance.
Inspection Report# : 2004012(pdf)
 
2Q/2005 Inspection Findings - River Bend 1                                                                                            Page 3 of 5 Mitigating Systems Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to inspect portable fire extinguishers within the required frequency The inspectors identified a Green noncited violation of Attachment 4 to Facility Operating License NPF-47 for failure to inspect portable fire extinguishers within the required frequency. The inspectors identified a total of 24 portable fire extinguishers that had not received an inspection during the month of April 2005. The inspectors found 28 condition reports in the licensee's corrective action program documenting missed inspections of portable fire extinguishers during the period from January 2000 through April 2005. Two of these condition reports were based on NRC-identified missed inspections of portable fire extinguishers in January and September of 2004.
The inspectors determined that this NRC-identified finding was more than minor because it was associated with the mitigating systems cornerstone attribute to protect against external factors, like fire, and because the finding affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The inspectors determined that the degradation rating was "low" because the fire extinguishers were expected to display nearly the same level of effectiveness and reliability as they would have had the fire extinguishers been inspected during the month of April 2005. Because this finding was assigned a low degradation rating, it was screened as having very low risk significance (Green). This finding also had crosscutting aspects associated with problem identification and resolution since the inspectors found 28 condition reports in the licensee's corrective action program documenting missed inspections of portable fire extinguishers during the period from January 2000 through April 2005. Because this Green finding was documented in the licensee's corrective action program as CR-RBS-2005-01726, this violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy.
Inspection Report# : 2005003(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Wide range reactor water level indication did not respond, as expected by operators, following an unplanned reactor scram A self-revealing, noncited violation of 10 CFR 55.46(c) was identified regarding differences between the simulator's and the plant's wide-range reactor water level digital indications during an unplanned reactor scram. This unexpected level indication resulted in indecision on the part of the operators during postscram recovery actions on December 10, 2004.
This finding is more than minor since deficiencies in the operator training program could become a more significant safety concern if left uncorrected. Based on the results of the significance determination process using Inspection Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process," this finding was determined to have very low safety significance, since it did not involve an exam or operating test but did involve a simulator fidelity issue which impacted operator actions during the response to an actual transient in the plant.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Rainwater leaked from auxiliary building roof onto Division I auxiliary building 480 Vac ESF switchgear, causing loss of a safety-related auxiliary building area unit The inspectors identified a self-revealing noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, for the licensee's failure to take timely and effective corrective action to prevent recurrence of rainwater leakage from the auxiliary building roof onto auxiliary building 480 Vac safety-related Switchgear EJS-SWGR2A, causing a loss of auxiliary building area unit Cooler HVR-UC11A. Investigation into the source of water determined that rainwater was accumulating inside the auxiliary building fresh air intake structure on the roof and leaking through seals along the air inlet ductwork onto Switchgear EJS-SWGR2A. The inspectors determined that this was a repeat of a February 5, 2004, leak documented in River Bend Station Condition Report 2004-0346 and a problem identification and resolution Noncited Violation 05000458/2004002-02. This finding had crosscutting aspects related to ineffective corrective actions.
The inspectors determined that the licensee's failure to take timely and effective corrective action to stop rainwater leaks from the auxiliary building roof onto Switchgear EJS-SWG2A was a performance deficiency that caused the loss of Cooler HVR-UC11A. The finding was more than minor because, if left uncorrected, rainwater leaks from the auxiliary building roof could lead to the loss of other Division I safety-related equipment and motor control centers powered by Switchgear EJS-SWG2A. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance because the short-term loss of unit Cooler HVR-UC11A did not cause an actual loss of safety function of any train of Technical Specification risk significant equipment and was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the failure to take timely and effective actions to prevent rainwater from leaking onto Switchgear EJS-SWGR2A was a violation of 10 CFR Part 50,
 
2Q/2005 Inspection Findings - River Bend 1                                                                                              Page 4 of 5 Appendix B, Criterion XVI, "Corrective Action." Because this finding was of very low safety significance and was entered into the licensee's corrective action program as CR-RBS-2004-4218, this violation is being treated as a noncited violation, consistent with Section IV.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Oct 08, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to implement a required procedure for loss of main condenser vacuum/trip of circulating water pumps The inspectors identified a non-cited violation of Technical Specifications 5.4.1.a for the failure of the licensee to implement the Abnormal Operating Procedure AOP-0005, "Loss of Main Condenser Vacuum/Trip of Circulating Water Pump," following the loss of two of three operating circulating water pumps. Failure to implement this procedure contributed to the loss of condenser vacuum. This finding had cross-cutting aspects of human performance in that the operators did not implement the abnormal operating procedure as required. Additionally, this finding had cross-cutting aspects regarding problem identification and resolution in that a similar event had occurred over a month earlier, and no actions were taken to incorporate that operating experience into the operating procedures or process it through the corrective action program.
This finding is greater than minor because it is associated with human performance attribute of the mitigating system cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding actually led to the loss of main condenser vacuum and forced the operators to perform a reactor cool down through safety relief valves, reactor core isolation cooling and the suppression pool. This finding is of very low safety significance because it would only affect the plant during this particular situation of partial loss of offsite power and that all mitigating capability was maintained.
Inspection Report# : 2004012(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to control special processes such as welding in accordance with qualified welding procedures The inspector identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion IX, for failure to control special processes, such as welding, in accordance with qualified welding procedures as required. The finding was a human performance error for the failure to follow procedure. Criterion IX, Appendix B, of 10 CFR Part 50, "Control of Special Processes," requires in part that measures shall be established to assure that special processes, including welding, heat treating, and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
Contrary to the above, welding personnel failed to verify interpass temperature during welding activities on feedwater inlet check Valve B21-AOVF032, an ASME Class1 valve, in accordance with qualified welding procedures.
This finding was determined to be more than minor, through Inspection Manual Chapter 0612, Appendix B, in that it affected the barrier integrity cornerstone attribute of human performance, could have represented a more significant issue if left uncorrected, and there was a reasonable likelihood that the valve would have been returned to service if the inspector had not intervened. Based on the results of a significance determination process Phase 1 analysis, this finding had very low safety significance because it did not result in the loss of a barrier integrity function and has been entered into the licensee's corrective action program as Condition Report CR-RBS-2004-03395. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation
 
2Q/2005 Inspection Findings - River Bend 1                                                                                          Page 5 of 5 Failure to control a high radiation area in accordance with Technical Specification 5.7.3 The inspector reviewed a self-revealing noncited violation of Technical Specification 5.7.3 because the licensee failed to control a high radiation area with dose rates greater than 1,000 millirems per hour. On October 31, 2004, during maintenance activities on valves located on the 82-foot level of the drywell, three workers' electronic alarming dosimeters unexpectedly alarmed when they were exposed to unanticipated radiation levels of approximately 1,700 millirems per hour. Subsequent radiation surveys at the source of radiation around Valve RCS-V-3009 identified 6,000 millirems per hour on contact and 2,000 millirems per hour at 30 centimeters. The area was not barricaded, conspicuously posted, and did not have a flashing light activated as a warning device. The licensee determined that the three workers received 84, 85, and 95 millirems, respectively. This finding was entered into the licensee's corrective action program.
This finding is more than minor because it is associated with the Occupational Radiation Safety attribute of exposure control and affected the cornerstone objective, in that not controlling locked high radiation areas could increase personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance (Green) 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.
Inspection Report# : 2004005(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : August 24, 2005
 
3Q/2005 Inspection Findings - River Bend 1                                                                                              Page 1 of 5 River Bend 1 3Q/2005 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to revise a tagging boundary to support an emergent troubleshooting task resulted in a loss of protected division of offsite power and shutdown cooling The inspectors identified a green noncited violation of Technical Specification 5.4.1.a for failure to make a proper change to the tagging boundary around balance of plant Transformer RTX-XSR1F during Refueling Outage 12. This performance deficiency resulted in a trip signal, generated during troubleshooting the transformer sudden overpressure protection circuit, which caused the trip of switchyard Breakers OCB-20670 and OCB-20665. This resulted in the loss of offsite power to Division II engineered safety features Transformer RTX-XSR1D, causing a loss of shutdown cooling, a loss of alternate decay heat removal, containment isolations, and an automatic start of the Division II emergency diesel generator.
The inspectors determined that this human performance error was more than minor because it was associated with the initiating event cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the finding using IMC 0609, Appendix G, "Shutdown Operations Significance Determination Process,"
and determined that the loss of offsite power to Division II engineered safety features switchgear was of very low safety significance because there was no increased likelihood of a loss of reactor coolant system inventory, there was no loss of reactor water level instrumentation, there was no degradation of the licensee's ability to terminate a leak path or add water to the reactor when needed, nor was there any degradation of the licensee's ability to recover decay heat removal once it was lost. Because this human performance error was of very low safety significance (Green) and was documented in the licensee's corrective action program as CR-RBS-2003-03456, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation Human performance error causes a loss of offsite power to Division I ESF Switchgear and start of the Division I emergency diesel generator during Refueling Outage 12 The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a. that was of very low safety significance (Green). As a result, during preparation for Division I integrated emergency core cooling systems testing, a technician inadvertently made contact with the wrong terminal on an undervoltage relay which tripped the preferred offsite power feeder breaker for the Division I safety-related 4160 Vac switchgear and started the Division I emergency diesel generator.
The inspectors determined that the inadvertent contact of the wrong terminal on Division I was a performance deficiency and a human performance error. Also, ineffective and incomplete corrective actions for similar errors contributed to the performance deficiency. The finding was more than minor because it was associated with the initiating events cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions, namely a partial loss of offsite power. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significant Determination Process," Attachment 1, Checklist 7, "BWR Refueling Operations with RCS Level greater than 23 feet." The finding was of very low safety significance (Green) because it did not cause a loss of shutdown cooling and did not compromise the ac power guidelines that: (1) one qualified circuit of offsite power remain operable; (2) at least one emergency diesel generator remain operable; and (3) necessary portions of the ac electrical power distribution systems remain operable.
The inspectors determined that this human performance error with problem identification and resolution aspects was the result of a violation of Technical Specification 5.4.1.a. which states, in part, that procedures shall be implemented and maintained as recommended in NUREG 1.33, Revision 2, Appendix A. Section 9.e. refers to general procedures for the control of maintenance activities. The licensee failed to evaluate the applicability of error reduction techniques, such as "taping of adjacent leads/contact points," for the installation of jumpers during Division I integrated emergency core cooling system testing, Procedure STP-309-0603, in accordance with Procedure ADM-0023, "Conduct of Maintenance," Revision 17A, Section 8.5. In addition, the licensee failed to install banana jacks on terminals on the back of the undervoltage relay in the Division I safety-related 4160 Vac switchgear, which were jumpered during the performance of Procedure STP-309-0603, in accordance with Procedure EDS-EE-001, "Banana Jack Standard," Revision 3. Because the finding was of very low safety significance and was entered into the licensee's corrective action program as Condition Report CR-RBS-2004-3518, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
 
3Q/2005 Inspection Findings - River Bend 1                                                                                              Page 2 of 5 Significance:        Dec 31, 2004 Identified By: NRC Item Type: FIN Finding Automatic reactor scram during main turbine control valve testing due to control system malfunction The inspectors identified a finding based on the licensee's failure to adequately identify the root cause of the April 21, 2001, turbine trip and reactor scram so as to prevent recurrence. This failure resulted in a subsequent turbine trip and reactor scram on September 22, 2003.
The inspectors determined that the failure by the licensee to adequately identify the root cause of the April 21, 2001, event and to take effective corrective actions to prevent electrostatic arcing from affecting the primary and backup speed probes, was a performance deficiency. The inspectors determined that this performance deficiency led directly to the recurrence of the event on September 22, 2003. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affect mitigating equipment, and did not increase the likelihood of a fire or flood. This finding had problem identification and resolution crosscutting aspects regarding ineffective root cause determinations (evaluation). It was entered into the licensee's corrective action program as Condition Report CR-RBS-2003-3203.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: FIN Finding Failure to identify and properly evaluate deficient conditions related to switchyard breakers The inspectors identified a self-revealing finding of very low safety significance concerning the licensee's failure to identify a deficient condition due to preconditioned speed testing of station switchyard breakers and properly evaluate three similar failures of station switchyard breakers. As a result, three switchyard breakers opened slowly on August 15, 2004, and a transmission line ground fault that should have been isolated from the station switchyard remained connected to the main transformer long enough to cause a main generator lockout and reactor scram. Additionally, because slow breaker opening deenergized the north 230 kV bus, isolation of a coincident transmission line fault resulted in a loss of power to half of the balance of plant loads and the Division II engineered safety features switchboard.
This problem identification and resolution finding was more than minor because it was associated with the initiating events cornerstone objective to limit those events that upset plant stability and challenge a critical safety function during power operations. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Because the finding contributed to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, the finding required a Phase 2 analysis. The inspectors referred the results of the Phase 2 analysis to the regional senior reactor analyst for final determination of risk.
The senior reactor analyst performed a Phase 3 analysis of the event. The factors that contributed to the result of that analysis included: (1) the dominant sequence was a transient with a loss of power to a vital bus; (2) the consequences of the finding were bounded by a complete loss of offsite power; (3) the history of single slow switchyard breaker operation; (4) the design and layout of the station switchyard; and (4) the possibility of recovery from either a partial or complete loss of offsite power given the conditions that led to the events of August 15, 2004. The result was that the finding was of very low safety significance.
Inspection Report# : 2004005(pdf)
Significance:        Oct 08, 2004 Identified By: Self-Revealing Item Type: FIN Finding Failure to maintain circulating water cooling tower drift eliminators and to take timely corrective actions to address insulator arcing The inspectors documented a self-revealing finding for failure to adequately maintain the circulating water cooling tower drift eliminators which resulted in salt contamination of the insulators in the on-site transformer yard, and failure to take corrective actions when pre-established trigger points were reached regarding insulator arcing (corona). The resulting contamination and failure to clean the insulators caused ground faults on Reserve Station Service Line1 and main transformers, which resulted in the loss of the Division I off-site power and a reactor scram on October 1, 2004. This finding had crosscutting aspects related to problem identification and resolution in that corrective actions were not implemented in a timely manner to prevent a significant plant transient.
This finding is more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. A completed Phase 3 evaluation resulted in an incremental conditional core damage probability of 1.2E-
: 7. Therefore, the significance of the finding was determined to be of very low safety significance.
Inspection Report# : 2004012(pdf)
 
3Q/2005 Inspection Findings - River Bend 1                                                                                            Page 3 of 5 Mitigating Systems Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to inspect portable fire extinguishers within the required frequency The inspectors identified a Green noncited violation of Attachment 4 to Facility Operating License NPF-47 for failure to inspect portable fire extinguishers within the required frequency. The inspectors identified a total of 24 portable fire extinguishers that had not received an inspection during the month of April 2005. The inspectors found 28 condition reports in the licensee's corrective action program documenting missed inspections of portable fire extinguishers during the period from January 2000 through April 2005. Two of these condition reports were based on NRC-identified missed inspections of portable fire extinguishers in January and September of 2004.
The inspectors determined that this NRC-identified finding was more than minor because it was associated with the mitigating systems cornerstone attribute to protect against external factors, like fire, and because the finding affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The inspectors determined that the degradation rating was "low" because the fire extinguishers were expected to display nearly the same level of effectiveness and reliability as they would have had the fire extinguishers been inspected during the month of April 2005. Because this finding was assigned a low degradation rating, it was screened as having very low risk significance (Green). This finding also had crosscutting aspects associated with problem identification and resolution since the inspectors found 28 condition reports in the licensee's corrective action program documenting missed inspections of portable fire extinguishers during the period from January 2000 through April 2005. Because this Green finding was documented in the licensee's corrective action program as CR-RBS-2005-01726, this violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy.
Inspection Report# : 2005003(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation Wide range reactor water level indication did not respond, as expected by operators, following an unplanned reactor scram A self-revealing, noncited violation of 10 CFR 55.46(c) was identified regarding differences between the simulator's and the plant's wide-range reactor water level digital indications during an unplanned reactor scram. This unexpected level indication resulted in indecision on the part of the operators during postscram recovery actions on December 10, 2004.
This finding is more than minor since deficiencies in the operator training program could become a more significant safety concern if left uncorrected. Based on the results of the significance determination process using Inspection Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process," this finding was determined to have very low safety significance, since it did not involve an exam or operating test but did involve a simulator fidelity issue which impacted operator actions during the response to an actual transient in the plant.
Inspection Report# : 2004005(pdf)
Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation Rainwater leaked from auxiliary building roof onto Division I auxiliary building 480 Vac ESF switchgear, causing loss of a safety-related auxiliary building area unit The inspectors identified a self-revealing noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, for the licensee's failure to take timely and effective corrective action to prevent recurrence of rainwater leakage from the auxiliary building roof onto auxiliary building 480 Vac safety-related Switchgear EJS-SWGR2A, causing a loss of auxiliary building area unit Cooler HVR-UC11A. Investigation into the source of water determined that rainwater was accumulating inside the auxiliary building fresh air intake structure on the roof and leaking through seals along the air inlet ductwork onto Switchgear EJS-SWGR2A. The inspectors determined that this was a repeat of a February 5, 2004, leak documented in River Bend Station Condition Report 2004-0346 and a problem identification and resolution Noncited Violation 05000458/2004002-02. This finding had crosscutting aspects related to ineffective corrective actions.
The inspectors determined that the licensee's failure to take timely and effective corrective action to stop rainwater leaks from the auxiliary building roof onto Switchgear EJS-SWG2A was a performance deficiency that caused the loss of Cooler HVR-UC11A. The finding was more than minor because, if left uncorrected, rainwater leaks from the auxiliary building roof could lead to the loss of other Division I safety-related equipment and motor control centers powered by Switchgear EJS-SWG2A. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." Based on the results of the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance because the short-term loss of unit Cooler HVR-UC11A did not cause an actual loss of safety function of any train of Technical Specification risk significant equipment and was not potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the failure to take timely and effective actions to prevent rainwater from leaking onto Switchgear EJS-SWGR2A was a violation of 10 CFR Part 50,
 
3Q/2005 Inspection Findings - River Bend 1                                                                                              Page 4 of 5 Appendix B, Criterion XVI, "Corrective Action." Because this finding was of very low safety significance and was entered into the licensee's corrective action program as CR-RBS-2004-4218, this violation is being treated as a noncited violation, consistent with Section IV.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Significance:        Oct 08, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to implement a required procedure for loss of main condenser vacuum/trip of circulating water pumps The inspectors identified a non-cited violation of Technical Specifications 5.4.1.a for the failure of the licensee to implement the Abnormal Operating Procedure AOP-0005, "Loss of Main Condenser Vacuum/Trip of Circulating Water Pump," following the loss of two of three operating circulating water pumps. Failure to implement this procedure contributed to the loss of condenser vacuum. This finding had cross-cutting aspects of human performance in that the operators did not implement the abnormal operating procedure as required. Additionally, this finding had cross-cutting aspects regarding problem identification and resolution in that a similar event had occurred over a month earlier, and no actions were taken to incorporate that operating experience into the operating procedures or process it through the corrective action program.
This finding is greater than minor because it is associated with human performance attribute of the mitigating system cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding actually led to the loss of main condenser vacuum and forced the operators to perform a reactor cool down through safety relief valves, reactor core isolation cooling and the suppression pool. This finding is of very low safety significance because it would only affect the plant during this particular situation of partial loss of offsite power and that all mitigating capability was maintained.
Inspection Report# : 2004012(pdf)
Barrier Integrity Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to control special processes such as welding in accordance with qualified welding procedures The inspector identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion IX, for failure to control special processes, such as welding, in accordance with qualified welding procedures as required. The finding was a human performance error for the failure to follow procedure. Criterion IX, Appendix B, of 10 CFR Part 50, "Control of Special Processes," requires in part that measures shall be established to assure that special processes, including welding, heat treating, and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
Contrary to the above, welding personnel failed to verify interpass temperature during welding activities on feedwater inlet check Valve B21-AOVF032, an ASME Class1 valve, in accordance with qualified welding procedures.
This finding was determined to be more than minor, through Inspection Manual Chapter 0612, Appendix B, in that it affected the barrier integrity cornerstone attribute of human performance, could have represented a more significant issue if left uncorrected, and there was a reasonable likelihood that the valve would have been returned to service if the inspector had not intervened. Based on the results of a significance determination process Phase 1 analysis, this finding had very low safety significance because it did not result in the loss of a barrier integrity function and has been entered into the licensee's corrective action program as Condition Report CR-RBS-2004-03395. This violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600.
Inspection Report# : 2004005(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation
 
3Q/2005 Inspection Findings - River Bend 1                                                                                          Page 5 of 5 Failure to control a high radiation area in accordance with Technical Specification 5.7.3 The inspector reviewed a self-revealing noncited violation of Technical Specification 5.7.3 because the licensee failed to control a high radiation area with dose rates greater than 1,000 millirems per hour. On October 31, 2004, during maintenance activities on valves located on the 82-foot level of the drywell, three workers' electronic alarming dosimeters unexpectedly alarmed when they were exposed to unanticipated radiation levels of approximately 1,700 millirems per hour. Subsequent radiation surveys at the source of radiation around Valve RCS-V-3009 identified 6,000 millirems per hour on contact and 2,000 millirems per hour at 30 centimeters. The area was not barricaded, conspicuously posted, and did not have a flashing light activated as a warning device. The licensee determined that the three workers received 84, 85, and 95 millirems, respectively. This finding was entered into the licensee's corrective action program.
This finding is more than minor because it is associated with the Occupational Radiation Safety attribute of exposure control and affected the cornerstone objective, in that not controlling locked high radiation areas could increase personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance (Green) 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.
Inspection Report# : 2004005(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Sep 09, 2005 Identified By: NRC Item Type: FIN Finding Unplanned Scrams Exceed the Criteria for a White Performance Indicator The U.S. Nuclear Regulatory Commission performed this supplemental inspection to assess the licensee's evaluations associated with four unplanned reactor scrams that occurred between August 15, 2004 and January 15, 2005. The cumulative effect of these trips was that the Performance Indicator for unplanned scrams per 7000 critical hours crossed the threshold from Green (very low risk significance) to White (low to moderate risk significance) for the first quarter of calendar year 2005. The licensee performed individual root cause evaluations for all of the four reactor scrams. In addition to the individual trip evaluations, the licensee performed a common cause analysis to identify any performance and process issues that led to the White performance indicator. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that for each scram, the licensee performed a comprehensive and thorough evaluation in which specific problems were identified, an adequate root cause evaluation was performed, and corrective actions were taken or planned to prevent recurrence.
Inspection Report# : 2005012(pdf)
Last modified : November 30, 2005
 
4Q/2005 Inspection Findings - River Bend 1                                                                                            Page 1 of 4 River Bend 1 4Q/2005 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to complete TS required actions within allowed completion time The NRC identified a noncited violation of Technical Specification 3.4.1.A for the licensee's failure to shut down one reactor recirculation loop within 2 hours of determining that jet pump loop flow mismatch was greater than 5 percent while operating at greater than 70 percent of rated core flow. On October 31, 2005, the Reactor Recirculation Flow Control Valve B hydraulic power unit tripped because of a blown control power fuse, causing Flow Control Valve B to drift open. Operators throttled closed Flow Control Valve A to maintain reactor power at 100 percent, resulting in a jet pump loop flow mismatch of approximately 8.2 percent. The flow mismatch existed for 4.5 hours. The licensee entered this into their corrective action program as Condition Report CR-RBS-2006-00274.
The finding was more than minor because, if left uncorrected, it would become a more significant safety concern. Matched recirculation loop flows is an assumption used in the accident analysis for a loss of coolant accident resulting from a loop break. A flow mismatch could result in core response that is more severe than assumed in the accident analysis. The significance of this finding could not be evaluated using MC 0609, "Significance Determination Process." Based on management review, the finding was determined to be of very low safety significance based on the short duration of the flow mismatch, 4.5 hours, and the low likelihood of a loss of coolant accident during that time. The cause of this finding is related to the crosscutting element of human performance in that operators failed to implement Technical Specification requirements.
Inspection Report# : 2005005(pdf)
Mitigating Systems Significance:        Dec 31, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate design assumption results in RCIC turbine exhaust header filling with water following an automatic high water level shutdown A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensee's failure to address the worst case conditions in the sizing calculation for the reactor core isolation cooling turbine exhaust line vacuum breaker system as part of a plant modification to remove the internals of the reactor core isolation cooling turbine exhaust line check valve. As a result, on December 10, 2004, when the reactor core isolation cooling system was started and subsequently shutdown on high reactor water level following a scram and loss of feedwater, the turbine exhaust line filled with water from the suppression pool, causing the operators to consider the system unavailable and complicating their response to the event. The licensee entered this finding into their corrective action program as CR-RBS-2005-00724 and reinstalled the turbine exhaust line check valve internals in February 2005.
The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective to ensure the availability and reliability of the reactor core isolation cooling system, a system that responds to initiating events (loss of feedwater and station blackout), to prevent undesirable consequences. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because it represented a design deficiency that did not result in a loss of system function.
Inspection Report# : 2005005(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Corrective Actions in Response to a 10 CFR Part 21 Report The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for failure to implement corrective actions in response to a 10 CFR Part 21 Report. The corrective actions involved performing vendor-recommended magnetic particle inspections of emergency diesel generator cylinder liners to look for cracks. During a records review in August 2005, the inspectors identified that in April 1999, two cylinder liners from the Division I emergency diesel generator were replaced but the required magnetic particle testing inspections were not performed.
 
4Q/2005 Inspection Findings - River Bend 1                                                                                            Page 2 of 4 This finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency power to respond to initiating events to prevent undesirable consequences. Since the finding did not represent an actual loss of safety function for either of the emergency diesel generators, the finding was determined to be of very low safety significance using Phase 1 of the Significant Determination Process. This finding had crosscutting aspects associated with problem identification and resolution. The licensee entered this finding into their corrective action program as CR-RBS-2005-03400.
Inspection Report# : 2005004(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: FIN Finding Failure to Troubleshoot a Starting System Failure Caused Station Blackout Diesel Generator to Be Unavailable for 24 Hours Longer than Necessary The inspectors identified a finding associated with the licensee's failure to perform adequate troubleshooting of a problem with the station blackout diesel generator that resulted in the diesel generator being out of service for 24 hours longer than necessary. Licensee personnel focused on the suspected cause, the engine starter, and did not perform comprehensive troubleshooting to identify the actual cause of the failure.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute for equipment performance and the cornerstone objective to ensure the availability of a system that responds to initiating events to prevent undesirable consequences. During Phase 2 of the significance determination process for at power situations, the finding screened as having very low safety significance (Green), because the station blackout diesel generator was unavailable for less than three days and the other diesel generators were available. The finding had crosscutting aspects associated with problem identification and resolution based on the fact that licensee personnel failed to properly assess the starting system failure. This finding is entered in the licensee's corrective action program as CR-RBS- 2005-02897.
Inspection Report# : 2005004(pdf)
Significance:        Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Completely Close a Residual Heat Removal System Valve Resulted in Pumping Suppression Pool Water to Containment Upper Pool A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for a failure to follow procedures. During motor-operated valve stroke time testing of Residual Heat Removal to Upper Pool Fuel Pool Cooling Assist Valve E12-MOVF037A, an operator failed to follow procedures by not completely closing Valve E12-F037A. As a result, when Residual Heat Removal System A was later operated in suppression pool cooling mode, approximately 5,000 gallons of suppression pool level was pumped to the containment upper pool. The licensee took immediate corrective action to identify and close all motor-operated throttle valves and issued a standing order to ensure all motor-operated throttle valves were completely closed when operated from the main control room.
The finding was more than minor because, if left uncorrected, the failure to completely close motor-operated throttle valves could become a more significant safety concern. Using the significance determination process, the inspectors determined that the finding was of very low safety significance (Green) because it was not a design or qualification issue and it did not represent an actual loss of safety function of either residual heat removal System A or the suppression pool. The inspectors determined that this finding had human performance and problem identification and resolution crosscutting aspects. The failure to completely close Valve E12-F037A was a human performance error caused by a lack of understanding of the operation of motor-operated throttle valves and inadequate guidance in the test procedure. The inspectors also determined that a similar event involving the same valve occurred during the last refueling outage, and the licensee failed to identify and correct the underlying cause of the performance deficiency. Because this failure to comply with TS 5.4.1.a. was of very low safety significance and was entered in the licensee's corrective action program as CR-RBS-2005-02772, the inspectors determined that it was a noncited violation in accordance with Section VI. A of the NRC Enforcement Policy.
Inspection Report# : 2005004(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to inspect portable fire extinguishers within the required frequency The inspectors identified a Green noncited violation of Attachment 4 to Facility Operating License NPF-47 for failure to inspect portable fire extinguishers within the required frequency. The inspectors identified a total of 24 portable fire extinguishers that had not received an inspection during the month of April 2005. The inspectors found 28 condition reports in the licensee's corrective action program documenting missed inspections of portable fire extinguishers during the period from January 2000 through April 2005. Two of these condition reports were based on NRC-identified missed inspections of portable fire extinguishers in January and September of 2004.
The inspectors determined that this NRC-identified finding was more than minor because it was associated with the mitigating systems cornerstone attribute to protect against external factors, like fire, and because the finding affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors
 
4Q/2005 Inspection Findings - River Bend 1                                                                                          Page 3 of 4 evaluated the finding using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The inspectors determined that the degradation rating was "low" because the fire extinguishers were expected to display nearly the same level of effectiveness and reliability as they would have had the fire extinguishers been inspected during the month of April 2005. Because this finding was assigned a low degradation rating, it was screened as having very low risk significance (Green). This finding also had crosscutting aspects associated with problem identification and resolution since the inspectors found 28 condition reports in the licensee's corrective action program documenting missed inspections of portable fire extinguishers during the period from January 2000 through April 2005. Because this Green finding was documented in the licensee's corrective action program as CR-RBS-2005-01726, this violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy.
Inspection Report# : 2005003(pdf)
Barrier Integrity Emergency Preparedness Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for implementation of an EAL The NRC identified a noncited violation of 10 CFR Part 50, Appendix E, Section IV. B., as a result of inadequate procedures for the implementation of an emergency action level. The criteria in Procedure EIP-2-001, "Classification of Emergencies," Revision 12, for declaring an Alert emergency action level based on primary coolant leak rate, relied solely on a computer generated leakrate report that would not be valid under all conditions. The licensee entered this finding into their corrective action program as CR-RBS-2005-03078 and issued Standing Order 192, as an interim corrective action, to provide additional criteria to determine whether a primary coolant leak rate Alert emergency action level declaration was required.
The finding is more than minor because it is associated with the Emergency Preparedness Cornerstone attribute of procedural quality and affects the cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inadequate procedure could result in a failure to declare an Alert emergency classification when required. Using Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," this finding was determined to be of very low safety significance since it was a failure to comply with a regulatory requirement associated with a risk-significant planning standard that did not result in the loss or degradation of that risk-significant planning standard function.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Sep 09, 2005 Identified By: NRC Item Type: FIN Finding
 
4Q/2005 Inspection Findings - River Bend 1                                                                                          Page 4 of 4 Unplanned Scrams Exceed the Criteria for a White Performance Indicator The U.S. Nuclear Regulatory Commission performed this supplemental inspection to assess the licensee's evaluations associated with four unplanned reactor scrams that occurred between August 15, 2004 and January 15, 2005. The cumulative effect of these trips was that the Performance Indicator for unplanned scrams per 7000 critical hours crossed the threshold from Green (very low risk significance) to White (low to moderate risk significance) for the first quarter of calendar year 2005. The licensee performed individual root cause evaluations for all of the four reactor scrams. In addition to the individual trip evaluations, the licensee performed a common cause analysis to identify any performance and process issues that led to the White performance indicator. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that for each scram, the licensee performed a comprehensive and thorough evaluation in which specific problems were identified, an adequate root cause evaluation was performed, and corrective actions were taken or planned to prevent recurrence.
Inspection Report# : 2005012(pdf)
Last modified : March 03, 2006
 
1Q/2006 Inspection Findings - River Bend 1                                                                                            Page 1 of 6 River Bend 1 1Q/2006 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to complete TS required actions within allowed completion time The NRC identified a noncited violation of Technical Specification 3.4.1.A for the licensee's failure to shut down one reactor recirculation loop within 2 hours of determining that jet pump loop flow mismatch was greater than 5 percent while operating at greater than 70 percent of rated core flow. On October 31, 2005, the Reactor Recirculation Flow Control Valve B hydraulic power unit tripped because of a blown control power fuse, causing Flow Control Valve B to drift open. Operators throttled closed Flow Control Valve A to maintain reactor power at 100 percent, resulting in a jet pump loop flow mismatch of approximately 8.2 percent. The flow mismatch existed for 4.5 hours. The licensee entered this into their corrective action program as Condition Report CR-RBS-2006-00274.
The finding was more than minor because, if left uncorrected, it would become a more significant safety concern. Matched recirculation loop flows is an assumption used in the accident analysis for a loss of coolant accident resulting from a loop break. A flow mismatch could result in core response that is more severe than assumed in the accident analysis. The significance of this finding could not be evaluated using MC 0609, "Significance Determination Process." Based on management review, the finding was determined to be of very low safety significance based on the short duration of the flow mismatch, 4.5 hours, and the low likelihood of a loss of coolant accident during that time. The cause of this finding is related to the crosscutting element of human performance in that operators failed to implement Technical Specification requirements.
Inspection Report# : 2005005(pdf)
Mitigating Systems Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Installation of Incorrect Relief Valve Caused Leak in Standby Service Water System A self-revealing noncited violation of Technical Specifications Section 5.4.1.a. was identified for the failure of procurement engineers to specify the correct replacement relief valve in a repetitive maintenance task to periodically replace thermal relief valves in the standby service water system. As a result, an incorrect valve was installed in the system which, following a system pressure transient, failed to reseat, creating a 10 gpm leak from the system. The valve was replaced and the issue was entered into the licensee's corrective action program as CR-RBS-2006-1054.
The finding is more than minor because it would become more significant if left uncorrected in that additional makeup to the standby service water system would be required during a sustained loss of off-site power. The finding affected the mitigating system cornerstone. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance because it did not result in the loss of the standby service water system safety function. The cause of the finding is related to the crosscutting element of problem identification and resolution because the problem which led to the installation of the incorrect valve had been previously identified and corrective actions were not effective in preventing recurrence.
Inspection Report# : 2006002(pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate maintenance results in a drywell steam leak from Low Pressure Coolant Injection Train A Testable Check Valve A self-revealing noncited violation of Technical Specification Section 5.4.1.a, was identified for the failure of mechanical maintenance technicians to correctly reassemble Low Pressure Coolant Injection Testable Check Valve E12-AOVF041A during Refueling Outage 12. As a result, a steam leak from a valve flange caused a rise in drywell unidentified leakage. The issue was entered into the licensees corrective action program as CR-RBS-2006-00546 and the valve was repaired.
The finding is more than minor because it would have become a more significant safety concern if left uncorrected. The leakage would have continued to increase during the cycle, and it would have continued to have an adverse affect on indicated reactor vessel level. Using the
 
1Q/2006 Inspection Findings - River Bend 1                                                                                                Page 2 of 6 Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance because it did not result in a loss of the low pressure coolant injection system safety function and was not potentially risk significant due to seismic, flooding, or severe weather related initiating events. The finding had crosscutting aspects associated with human performance in that maintenance technicians incorrectly reassembled the valve during refueling outage 12.
Inspection Report# : 2006002(pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadvertent Initiation of High Pressure Core Spray Caused by the Use of the Wrong Test Plug During Surveillance Testing A self-revealing noncited violation of Technical Specifications Section 5.4.1.a. was identified for the failure to provide adequate procedural guidance for the use of a test plug during the performance of a required surveillance test procedure. The use of the wrong test plug caused an initiation of the high pressure core spray system and injection into the vessel. The issue was entered in the licensees corrective action program as CR-RBS-2006-00283.
The finding is more than minor because it is associated with the mitigating system cornerstone attribute of equipment performance and the cornerstone objective to ensure the availability and reliability of high pressure core spray, a system that responds to initiating events to prevent undesirable consequences. The Phase 1 worksheets in Manual Chapter 0609, "Significance Determination Process," were used to conclude that a Phase 2 analysis was required because there was an actual loss of system safety function. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding is related to the crosscutting element of human performance because the technicians did not verify that they were using the correct test plug for the surveillance test being performed.
Inspection Report# : 2006002(pdf)
Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable Preconditioning of a Safety-related Valve Prior to Surveillance Testing The team identified a non-cited violation of Technical Specification 5.4.1.a (Procedures) for unacceptable preconditioning of a low pressure core spray keepfill system check valve. The test procedure failed to prescribe testing the check valve in the as-found condition. Instead (during testing of the system pump) the document directed operators to flush the valve at 27 gpm for up to 20 minutes prior to the check valve test.
Corrosion buildup in the valve, which had previously caused valve failures, was a known concern and the preconditioning could have masked performance problems. Failure of the valve to perform its safety function puts the low pressure core spray system at risk of water hammer during a loss of offsite power event. The licensee planned to test the valve in the as-found configuration during future tests. The licensee documented this issue in their corrective action program as CR-RBS-2005-04123.
The failure to properly test the subject check valve was a performance deficiency. The finding was more than minor because, if left uncorrected, the problem could result in a more significant safety concern. Specifically, the surveillance test may not identify valve failure. The finding was of very low risk significance because it was not a design/qualification issue, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its technical specification allowable outage time, did not result in a loss of function of non safety-related risk significant equipment and was not risk significant due to external events. The finding had problem identification and resolution cross-cutting aspects because the licensee had failed to properly evaluate the issue as preconditioning in response to readily available industry information.
Inspection Report# : 2005008(pdf)
Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Replacement of a Valve to Correct a Significant Condition Adverse to Quality The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Actions) for the failure to take prompt corrective measures to address a significant condition adverse to quality. Specifically, the low pressure core spray keepfill pump discharge check valve failed on two occasions (significant conditions adverse to quality) and planned corrective measures to replace the check valve were not timely. The check valve failures put the low pressure core spray system at increased water hammer risk during a loss of offsite power event.
The licensee had identified that corrosion buildup was causing the valve to leak excessively when closed. The licensee documented this issue in their corrective action program as CR-RBS-2005-04162 and planned to replace the valve at the next available opportunity.
The failure to take prompt corrective measures to address a significant condition adverse to quality was a performance deficiency. The finding was greater than minor because it was an equipment performance reliability issue which impacted the mitigating systems cornerstone objective to ensure the reliability of systems that respond to initiating events. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the finding was of very low risk significance because it was not a design/qualification issue, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its technical specification allowable outage time, did not result in a loss of function of non safety-related risk significant equipment and was not risk significant due to external events. The finding had cross-cutting aspects in the area of problem identification and resolution.
Inspection Report# : 2005008(pdf)
 
1Q/2006 Inspection Findings - River Bend 1                                                                                            Page 3 of 6 Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Set MOV Limit Switches in Accordance with Design Documentation The team identified a 10 CFR 50, Appendix B, Criterion V (procedures) non-cited violation for the failure to set safety-related limit switches in accordance documents appropriate to the circumstances for 34 safety-related throttle valves. The licensee set motor-operated valve (MOV) open indication light limit switches so that the open indication de-energized between the 95% and 100% closed positions, whereas the applicable procedure and design drawing required that the limit switches be set to the 100% closed position. This practice had caused repetitive operational problems in the plant. The licensee entered this issue into their corrective action program as CR-RBS-2005-04113.
The failure to adjust MOV limit switches in accordance with documents appropriate to the circumstances was a performance deficiency. The issue was more than minor because it affected the mitigating systems cornerstone objective, in that it affected the operability, availability, reliability or function of a system or train in a mitigating system. The finding was of very low safety significance because it was a design/qualification deficiency confirmed not to result in loss of operability per "Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment." This finding had cross-cutting aspects in the areas of human performance, (the failure to follow procedures) and problem identification and resolution because the licensee failed to identify the problem in response to a prior related NRC violation.
Inspection Report# : 2005008(pdf)
Significance:        Dec 31, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate design assumption results in RCIC turbine exhaust header filling with water following an automatic high water level shutdown A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensee's failure to address the worst case conditions in the sizing calculation for the reactor core isolation cooling turbine exhaust line vacuum breaker system as part of a plant modification to remove the internals of the reactor core isolation cooling turbine exhaust line check valve. As a result, on December 10, 2004, when the reactor core isolation cooling system was started and subsequently shutdown on high reactor water level following a scram and loss of feedwater, the turbine exhaust line filled with water from the suppression pool, causing the operators to consider the system unavailable and complicating their response to the event. The licensee entered this finding into their corrective action program as CR-RBS-2005-00724 and reinstalled the turbine exhaust line check valve internals in February 2005.
The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective to ensure the availability and reliability of the reactor core isolation cooling system, a system that responds to initiating events (loss of feedwater and station blackout), to prevent undesirable consequences. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because it represented a design deficiency that did not result in a loss of system function.
Inspection Report# : 2005005(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Corrective Actions in Response to a 10 CFR Part 21 Report The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for failure to implement corrective actions in response to a 10 CFR Part 21 Report. The corrective actions involved performing vendor-recommended magnetic particle inspections of emergency diesel generator cylinder liners to look for cracks. During a records review in August 2005, the inspectors identified that in April 1999, two cylinder liners from the Division I emergency diesel generator were replaced but the required magnetic particle testing inspections were not performed.
This finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency power to respond to initiating events to prevent undesirable consequences. Since the finding did not represent an actual loss of safety function for either of the emergency diesel generators, the finding was determined to be of very low safety significance using Phase 1 of the Significant Determination Process. This finding had crosscutting aspects associated with problem identification and resolution. The licensee entered this finding into their corrective action program as CR-RBS-2005-03400.
Inspection Report# : 2005004(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: FIN Finding Failure to Troubleshoot a Starting System Failure Caused Station Blackout Diesel Generator to Be Unavailable for 24 Hours Longer than Necessary The inspectors identified a finding associated with the licensee's failure to perform adequate troubleshooting of a problem with the station
 
1Q/2006 Inspection Findings - River Bend 1                                                                                            Page 4 of 6 blackout diesel generator that resulted in the diesel generator being out of service for 24 hours longer than necessary. Licensee personnel focused on the suspected cause, the engine starter, and did not perform comprehensive troubleshooting to identify the actual cause of the failure.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute for equipment performance and the cornerstone objective to ensure the availability of a system that responds to initiating events to prevent undesirable consequences. During Phase 2 of the significance determination process for at power situations, the finding screened as having very low safety significance (Green), because the station blackout diesel generator was unavailable for less than three days and the other diesel generators were available. The finding had crosscutting aspects associated with problem identification and resolution based on the fact that licensee personnel failed to properly assess the starting system failure. This finding is entered in the licensee's corrective action program as CR-RBS- 2005-02897.
Inspection Report# : 2005004(pdf)
Significance:        Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Completely Close a Residual Heat Removal System Valve Resulted in Pumping Suppression Pool Water to Containment Upper Pool A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for a failure to follow procedures. During motor-operated valve stroke time testing of Residual Heat Removal to Upper Pool Fuel Pool Cooling Assist Valve E12-MOVF037A, an operator failed to follow procedures by not completely closing Valve E12-F037A. As a result, when Residual Heat Removal System A was later operated in suppression pool cooling mode, approximately 5,000 gallons of suppression pool level was pumped to the containment upper pool. The licensee took immediate corrective action to identify and close all motor-operated throttle valves and issued a standing order to ensure all motor-operated throttle valves were completely closed when operated from the main control room.
The finding was more than minor because, if left uncorrected, the failure to completely close motor-operated throttle valves could become a more significant safety concern. Using the significance determination process, the inspectors determined that the finding was of very low safety significance (Green) because it was not a design or qualification issue and it did not represent an actual loss of safety function of either residual heat removal System A or the suppression pool. The inspectors determined that this finding had human performance and problem identification and resolution crosscutting aspects. The failure to completely close Valve E12-F037A was a human performance error caused by a lack of understanding of the operation of motor-operated throttle valves and inadequate guidance in the test procedure. The inspectors also determined that a similar event involving the same valve occurred during the last refueling outage, and the licensee failed to identify and correct the underlying cause of the performance deficiency. Because this failure to comply with TS 5.4.1.a. was of very low safety significance and was entered in the licensee's corrective action program as CR-RBS-2005-02772, the inspectors determined that it was a noncited violation in accordance with Section VI. A of the NRC Enforcement Policy.
Inspection Report# : 2005004(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to inspect portable fire extinguishers within the required frequency The inspectors identified a Green noncited violation of Attachment 4 to Facility Operating License NPF-47 for failure to inspect portable fire extinguishers within the required frequency. The inspectors identified a total of 24 portable fire extinguishers that had not received an inspection during the month of April 2005. The inspectors found 28 condition reports in the licensee's corrective action program documenting missed inspections of portable fire extinguishers during the period from January 2000 through April 2005. Two of these condition reports were based on NRC-identified missed inspections of portable fire extinguishers in January and September of 2004.
The inspectors determined that this NRC-identified finding was more than minor because it was associated with the mitigating systems cornerstone attribute to protect against external factors, like fire, and because the finding affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The inspectors determined that the degradation rating was "low" because the fire extinguishers were expected to display nearly the same level of effectiveness and reliability as they would have had the fire extinguishers been inspected during the month of April 2005. Because this finding was assigned a low degradation rating, it was screened as having very low risk significance (Green). This finding also had crosscutting aspects associated with problem identification and resolution since the inspectors found 28 condition reports in the licensee's corrective action program documenting missed inspections of portable fire extinguishers during the period from January 2000 through April 2005. Because this Green finding was documented in the licensee's corrective action program as CR-RBS-2005-01726, this violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy.
Inspection Report# : 2005003(pdf)
Barrier Integrity
 
1Q/2006 Inspection Findings - River Bend 1                                                                                            Page 5 of 6 Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain MCPR within Operating Limits The team identified two examples of a Technical Specification 3.2.2, "Minimum Critical Power Ratio" (MCPR), non-cited violation for the failure to prevent transition boiling on the fuel during Operational Cycles 8 and 11. Fuel failures due to transition boiling were experienced during each cycle. Engineers failed to properly understand the affect of zinc injection on the cladding surfaces following the Cycle 8 fuel pin failures and zinc injection was reinitiated before the corrective actions to prevent recurrence were in place. The licensee had industry information that indicated that zinc injection contributed to the accumulation of loose crud and the formation of tenacious crud on the fuel. The additional crud rendered the Technical Specifications Minimum Critical Power Ratio (MCPR) calculations inaccurate and transition boiling occurred in localized areas. The licensee entered this issue into their corrective action program as CR-RBS-2006-0255.
The failure to prevent transition boiling in the core was a performance deficiency. The issue was more than minor because it impacted the barrier integrity cornerstone objective to maintain the integrity of the fuel cladding. The finding screened out as of very low safety significance (Green) because it only affected the fuel barrier. The issue had cross-cutting aspects in the areas of problem identification and resolution, in that the licensee failed to properly evaluate pertinent related industry information, which could have precluded the first violation, and failed to properly implement effective corrective measures in response to the first set of fuel failures, which led to the second violation.
Inspection Report# : 2005008(pdf)
Emergency Preparedness Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for implementation of an EAL The NRC identified a noncited violation of 10 CFR Part 50, Appendix E, Section IV. B., as a result of inadequate procedures for the implementation of an emergency action level. The criteria in Procedure EIP-2-001, "Classification of Emergencies," Revision 12, for declaring an Alert emergency action level based on primary coolant leak rate, relied solely on a computer generated leakrate report that would not be valid under all conditions. The licensee entered this finding into their corrective action program as CR-RBS-2005-03078 and issued Standing Order 192, as an interim corrective action, to provide additional criteria to determine whether a primary coolant leak rate Alert emergency action level declaration was required.
The finding is more than minor because it is associated with the Emergency Preparedness Cornerstone attribute of procedural quality and affects the cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inadequate procedure could result in a failure to declare an Alert emergency classification when required. Using Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," this finding was determined to be of very low safety significance since it was a failure to comply with a regulatory requirement associated with a risk-significant planning standard that did not result in the loss or degradation of that risk-significant planning standard function.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
 
1Q/2006 Inspection Findings - River Bend 1                                                                                            Page 6 of 6 Miscellaneous Significance: N/A Jan 19, 2006 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Biannual Assessment The team reviewed approximately 225 condition reports, apparent and root cause analyses, as well as supporting documents to assess problem identification and resolution activities. In general, the corrective action program procedures and processes were effective, thresholds for identifying issues were low, and corrective actions were adequate to address conditions adverse to quality. Notwithstanding the above, poor engineering rigor associated with the prioritization and evaluation of issues resulted in a relatively high number of self-revealing and NRC identified findings. Some of these findings culminated in plant scrams and/or complicated operator response to emergency events. Others were related to equipment deficiencies, some of which resulted in inoperable safety-related equipment.
Based on the interviews conducted, the team concluded that a positive safety conscious work environment exists at River Bend Station. The team determined that employees felt free to raise safety concerns to station managers and supervisors, the employee concerns program, and the NRC. However, the team received a few isolated comments regarding the correction action program feedback process. These individuals had previously identified corrective action issues and were not satisfied with the program's responses to their concerns. Some of these individuals commented that they were hesitant to use the corrective action program in the future. The licensee acknowledged the comments and planned to take action to address the concerns. All the interviewees believed that potential safety issues were being addressed.
Inspection Report# : 2005008(pdf)
Significance: N/A Sep 09, 2005 Identified By: NRC Item Type: FIN Finding Unplanned Scrams Exceed the Criteria for a White Performance Indicator The U.S. Nuclear Regulatory Commission performed this supplemental inspection to assess the licensee's evaluations associated with four unplanned reactor scrams that occurred between August 15, 2004 and January 15, 2005. The cumulative effect of these trips was that the Performance Indicator for unplanned scrams per 7000 critical hours crossed the threshold from Green (very low risk significance) to White (low to moderate risk significance) for the first quarter of calendar year 2005. The licensee performed individual root cause evaluations for all of the four reactor scrams. In addition to the individual trip evaluations, the licensee performed a common cause analysis to identify any performance and process issues that led to the White performance indicator. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that for each scram, the licensee performed a comprehensive and thorough evaluation in which specific problems were identified, an adequate root cause evaluation was performed, and corrective actions were taken or planned to prevent recurrence.
Inspection Report# : 2005012(pdf)
Last modified : May 25, 2006
 
2Q/2006 Inspection Findings - River Bend 1                                                                                                    Page 1 of 7 River Bend 1 2Q/2006 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to complete TS required actions within allowed completion time The NRC identified a noncited violation of Technical Specification 3.4.1.A for the licensee's failure to shut down one reactor recirculation loop within 2 hours of determining that jet pump loop flow mismatch was greater than 5 percent while operating at greater than 70 percent of rated core flow. On October 31, 2005, the Reactor Recirculation Flow Control Valve B hydraulic power unit tripped because of a blown control power fuse, causing Flow Control Valve B to drift open. Operators throttled closed Flow Control Valve A to maintain reactor power at 100 percent, resulting in a jet pump loop flow mismatch of approximately 8.2 percent. The flow mismatch existed for 4.5 hours. The licensee entered this into their corrective action program as Condition Report CR-RBS-2006-00274.
The finding was more than minor because, if left uncorrected, it would become a more significant safety concern. Matched recirculation loop flows is an assumption used in the accident analysis for a loss of coolant accident resulting from a loop break. A flow mismatch could result in core response that is more severe than assumed in the accident analysis. The significance of this finding could not be evaluated using MC 0609, "Significance Determination Process." Based on management review, the finding was determined to be of very low safety significance based on the short duration of the flow mismatch, 4.5 hours, and the low likelihood of a loss of coolant accident during that time. The cause of this finding is related to the crosscutting element of human performance in that operators failed to implement Technical Specification requirements.
Inspection Report# : 2005005(pdf)
Mitigating Systems Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to identify Division III ESF bus supply breaker not racked in A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was reviewed involving the failure of the licensee to identify that the normal supply breaker to the Division III 4.16 kV engineered safety features bus was not properly racked in for a period of 24 days following maintenance. This issue was entered into the licensee's corrective action program as CR-RBS-2006-02402.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of configuration control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609, "Significance Determination Process," a Phase 3 analysis concluded that the finding was of very low safety significance. The cause of the finding was related to the crosscutting aspect of problem identification and resolution in that the licensee failed to properly evaluate available indications to identify that the breaker was not properly racked in.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately manage an increase in plant risk An NRC identified noncited violation of 10 CFR 50.65 Maintenance Rule Section (a)(4) was identified for the failure of the licensee to provide prescribed compensatory measures for two Orange shutdown risk conditions during Refueling Outage 13. Specifically, the preoutage risk assessment recommended that two work orders be in place for maintenance electricians to provide power to one spent fuel pool cooling pump in the event of problems with the running pump during periods of electrical bus maintenance. The inspectors found that the work packages were not in place before entering shutdown risk condition Orange on April 26, 2006, during the Division II engineering safety features bus testing, and May 3, 2006, during the Division I engineered safety features bus outage. This issue was entered into the licensee's corrective action program as CR-RBS-2006-01937.
The finding was more than minor because the licensee failed to implement a prescribed compensatory measure during the highest risk condition of Refueling Outage 13. The specific compensatory measures were called for in the preoutage risk assessment and the shutdown operations protection plan. The finding affected the mitigating system cornerstone because of the increased risk of a sustained loss of spent fuel pool cooling during core
 
2Q/2006 Inspection Findings - River Bend 1                                                                                                    Page 2 of 7 offloading operations. The finding could not be evaluated using the significance determination process, therefore the finding was reviewed by regional management and determined to be of very low safety significance. Factors that were considered included: (1) electrical maintenance technicians had previously performed the task of providing alternate power to a spent fuel pool cooling pump, (2) the necessary equipment was staged as part of the abnormal operating procedure for loss of decay heat removal, and (3) the relatively long time to boil of the spent fuel storage pool at that time during the refueling outage. The cause of the finding was related to the crosscutting aspect of human performance because the licensees planned maintenance activities and the predetermined increase in outage risk was not effectively managed by prescribed compensatory measures.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure to verify required offsite power breaker alignment An NRC identified noncited violation of Technical Specification 5.4.1.a was identified for the failure of the licensee to provide an adequate surveillance test procedure to perform Technical Specification Surveillance Requirement 3.8.1.1. Specifically, STP-000-0102, Power Distribution Alignment Check, Revision 4, did not verify the required offsite power circuit breaker alignment and indicated power availability for the Division III 4.16 kV engineered safety features bus as required in Modes 1, 2, and 3. This issue was entered into the licensee's corrective action program as CR-RBS-2006-02675 and -02402.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of configuration control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609, "Significance Determination Process," a Phase 3 analysis concluded that the finding was of very low safety significance.
Inspection Report# : 2006003(pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Installation of Incorrect Relief Valve Caused Leak in Standby Service Water System A self-revealing noncited violation of Technical Specifications Section 5.4.1.a. was identified for the failure of procurement engineers to specify the correct replacement relief valve in a repetitive maintenance task to periodically replace thermal relief valves in the standby service water system. As a result, an incorrect valve was installed in the system which, following a system pressure transient, failed to reseat, creating a 10 gpm leak from the system. The valve was replaced and the issue was entered into the licensee's corrective action program as CR-RBS-2006-1054.
The finding is more than minor because it would become more significant if left uncorrected in that additional makeup to the standby service water system would be required during a sustained loss of off-site power. The finding affected the mitigating system cornerstone. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance because it did not result in the loss of the standby service water system safety function. The cause of the finding is related to the crosscutting element of problem identification and resolution because the problem which led to the installation of the incorrect valve had been previously identified and corrective actions were not effective in preventing recurrence.
Inspection Report# : 2006002(pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate maintenance results in a drywell steam leak from Low Pressure Coolant Injection Train A Testable Check Valve A self-revealing noncited violation of Technical Specification Section 5.4.1.a, was identified for the failure of mechanical maintenance technicians to correctly reassemble Low Pressure Coolant Injection Testable Check Valve E12-AOVF041A during Refueling Outage 12. As a result, a steam leak from a valve flange caused a rise in drywell unidentified leakage. The issue was entered into the licensees corrective action program as CR-RBS-2006-00546 and the valve was repaired.
The finding is more than minor because it would have become a more significant safety concern if left uncorrected. The leakage would have continued to increase during the cycle, and it would have continued to have an adverse affect on indicated reactor vessel level. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance because it did not result in a loss of the low pressure coolant injection system safety function and was not potentially risk significant due to seismic, flooding, or severe weather related initiating events. The finding had crosscutting aspects associated with human performance in that maintenance technicians incorrectly reassembled the valve during refueling outage 12.
Inspection Report# : 2006002(pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation
 
2Q/2006 Inspection Findings - River Bend 1                                                                                                      Page 3 of 7 Inadvertent Initiation of High Pressure Core Spray Caused by the Use of the Wrong Test Plug During Surveillance Testing A self-revealing noncited violation of Technical Specifications Section 5.4.1.a. was identified for the failure to provide adequate procedural guidance for the use of a test plug during the performance of a required surveillance test procedure. The use of the wrong test plug caused an initiation of the high pressure core spray system and injection into the vessel. The issue was entered in the licensees corrective action program as CR-RBS-2006-00283.
The finding is more than minor because it is associated with the mitigating system cornerstone attribute of equipment performance and the cornerstone objective to ensure the availability and reliability of high pressure core spray, a system that responds to initiating events to prevent undesirable consequences. The Phase 1 worksheets in Manual Chapter 0609, "Significance Determination Process," were used to conclude that a Phase 2 analysis was required because there was an actual loss of system safety function. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding is related to the crosscutting element of human performance because the technicians did not verify that they were using the correct test plug for the surveillance test being performed.
Inspection Report# : 2006002(pdf)
Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable Preconditioning of a Safety-related Valve Prior to Surveillance Testing The team identified a non-cited violation of Technical Specification 5.4.1.a (Procedures) for unacceptable preconditioning of a low pressure core spray keepfill system check valve. The test procedure failed to prescribe testing the check valve in the as-found condition. Instead (during testing of the system pump) the document directed operators to flush the valve at 27 gpm for up to 20 minutes prior to the check valve test. Corrosion buildup in the valve, which had previously caused valve failures, was a known concern and the preconditioning could have masked performance problems.
Failure of the valve to perform its safety function puts the low pressure core spray system at risk of water hammer during a loss of offsite power event. The licensee planned to test the valve in the as-found configuration during future tests. The licensee documented this issue in their corrective action program as CR-RBS-2005-04123.
The failure to properly test the subject check valve was a performance deficiency. The finding was more than minor because, if left uncorrected, the problem could result in a more significant safety concern. Specifically, the surveillance test may not identify valve failure. The finding was of very low risk significance because it was not a design/qualification issue, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its technical specification allowable outage time, did not result in a loss of function of non safety-related risk significant equipment and was not risk significant due to external events. The finding had problem identification and resolution cross-cutting aspects because the licensee had failed to properly evaluate the issue as preconditioning in response to readily available industry information.
Inspection Report# : 2005008(pdf)
Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Replacement of a Valve to Correct a Significant Condition Adverse to Quality The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Actions) for the failure to take prompt corrective measures to address a significant condition adverse to quality. Specifically, the low pressure core spray keepfill pump discharge check valve failed on two occasions (significant conditions adverse to quality) and planned corrective measures to replace the check valve were not timely. The check valve failures put the low pressure core spray system at increased water hammer risk during a loss of offsite power event. The licensee had identified that corrosion buildup was causing the valve to leak excessively when closed. The licensee documented this issue in their corrective action program as CR-RBS-2005-04162 and planned to replace the valve at the next available opportunity.
The failure to take prompt corrective measures to address a significant condition adverse to quality was a performance deficiency. The finding was greater than minor because it was an equipment performance reliability issue which impacted the mitigating systems cornerstone objective to ensure the reliability of systems that respond to initiating events. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the finding was of very low risk significance because it was not a design/qualification issue, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its technical specification allowable outage time, did not result in a loss of function of non safety-related risk significant equipment and was not risk significant due to external events. The finding had cross-cutting aspects in the area of problem identification and resolution.
Inspection Report# : 2005008(pdf)
Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Set MOV Limit Switches in Accordance with Design Documentation The team identified a 10 CFR 50, Appendix B, Criterion V (procedures) non-cited violation for the failure to set safety-related limit switches in accordance documents appropriate to the circumstances for 34 safety-related throttle valves. The licensee set motor-operated valve (MOV) open indication light limit switches so that the open indication de-energized between the 95% and 100% closed positions, whereas the applicable procedure and design drawing required that the limit switches be set to the 100% closed position. This practice had caused repetitive operational problems in the plant. The licensee entered this issue into their corrective action program as CR-RBS-2005-04113.
 
2Q/2006 Inspection Findings - River Bend 1                                                                                                  Page 4 of 7 The failure to adjust MOV limit switches in accordance with documents appropriate to the circumstances was a performance deficiency. The issue was more than minor because it affected the mitigating systems cornerstone objective, in that it affected the operability, availability, reliability or function of a system or train in a mitigating system. The finding was of very low safety significance because it was a design/qualification deficiency confirmed not to result in loss of operability per "Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment." This finding had cross-cutting aspects in the areas of human performance, (the failure to follow procedures) and problem identification and resolution because the licensee failed to identify the problem in response to a prior related NRC violation.
Inspection Report# : 2005008(pdf)
Significance:        Dec 31, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate design assumption results in RCIC turbine exhaust header filling with water following an automatic high water level shutdown A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensee's failure to address the worst case conditions in the sizing calculation for the reactor core isolation cooling turbine exhaust line vacuum breaker system as part of a plant modification to remove the internals of the reactor core isolation cooling turbine exhaust line check valve. As a result, on December 10, 2004, when the reactor core isolation cooling system was started and subsequently shutdown on high reactor water level following a scram and loss of feedwater, the turbine exhaust line filled with water from the suppression pool, causing the operators to consider the system unavailable and complicating their response to the event. The licensee entered this finding into their corrective action program as CR-RBS-2005-00724 and reinstalled the turbine exhaust line check valve internals in February 2005.
The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective to ensure the availability and reliability of the reactor core isolation cooling system, a system that responds to initiating events (loss of feedwater and station blackout), to prevent undesirable consequences. Using Manual Chapter 0609, "Significance Determination Process,"
Phase 1 Worksheet, the finding was determined to have very low safety significance because it represented a design deficiency that did not result in a loss of system function.
Inspection Report# : 2005005(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Corrective Actions in Response to a 10 CFR Part 21 Report The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for failure to implement corrective actions in response to a 10 CFR Part 21 Report. The corrective actions involved performing vendor-recommended magnetic particle inspections of emergency diesel generator cylinder liners to look for cracks. During a records review in August 2005, the inspectors identified that in April 1999, two cylinder liners from the Division I emergency diesel generator were replaced but the required magnetic particle testing inspections were not performed.
This finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the capability of emergency power to respond to initiating events to prevent undesirable consequences. Since the finding did not represent an actual loss of safety function for either of the emergency diesel generators, the finding was determined to be of very low safety significance using Phase 1 of the Significant Determination Process. This finding had crosscutting aspects associated with problem identification and resolution. The licensee entered this finding into their corrective action program as CR-RBS-2005-03400.
Inspection Report# : 2005004(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: FIN Finding Failure to Troubleshoot a Starting System Failure Caused Station Blackout Diesel Generator to Be Unavailable for 24 Hours Longer than Necessary The inspectors identified a finding associated with the licensee's failure to perform adequate troubleshooting of a problem with the station blackout diesel generator that resulted in the diesel generator being out of service for 24 hours longer than necessary. Licensee personnel focused on the suspected cause, the engine starter, and did not perform comprehensive troubleshooting to identify the actual cause of the failure.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute for equipment performance and the cornerstone objective to ensure the availability of a system that responds to initiating events to prevent undesirable consequences. During Phase 2 of the significance determination process for at power situations, the finding screened as having very low safety significance (Green), because the station blackout diesel generator was unavailable for less than three days and the other diesel generators were available. The finding had crosscutting aspects associated with problem identification and resolution based on the fact that licensee personnel failed to properly assess the starting system failure. This finding is entered in the licensee's corrective action program as CR-RBS- 2005-02897.
Inspection Report# : 2005004(pdf)
Significance:        Sep 30, 2005 Identified By: Self-Revealing
 
2Q/2006 Inspection Findings - River Bend 1                                                                                                  Page 5 of 7 Item Type: NCV NonCited Violation Failure to Completely Close a Residual Heat Removal System Valve Resulted in Pumping Suppression Pool Water to Containment Upper Pool A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for a failure to follow procedures. During motor-operated valve stroke time testing of Residual Heat Removal to Upper Pool Fuel Pool Cooling Assist Valve E12-MOVF037A, an operator failed to follow procedures by not completely closing Valve E12-F037A. As a result, when Residual Heat Removal System A was later operated in suppression pool cooling mode, approximately 5,000 gallons of suppression pool level was pumped to the containment upper pool. The licensee took immediate corrective action to identify and close all motor-operated throttle valves and issued a standing order to ensure all motor-operated throttle valves were completely closed when operated from the main control room.
The finding was more than minor because, if left uncorrected, the failure to completely close motor-operated throttle valves could become a more significant safety concern. Using the significance determination process, the inspectors determined that the finding was of very low safety significance (Green) because it was not a design or qualification issue and it did not represent an actual loss of safety function of either residual heat removal System A or the suppression pool. The inspectors determined that this finding had human performance and problem identification and resolution crosscutting aspects. The failure to completely close Valve E12-F037A was a human performance error caused by a lack of understanding of the operation of motor-operated throttle valves and inadequate guidance in the test procedure. The inspectors also determined that a similar event involving the same valve occurred during the last refueling outage, and the licensee failed to identify and correct the underlying cause of the performance deficiency. Because this failure to comply with TS 5.4.1.a. was of very low safety significance and was entered in the licensee's corrective action program as CR-RBS-2005-02772, the inspectors determined that it was a noncited violation in accordance with Section VI. A of the NRC Enforcement Policy.
Inspection Report# : 2005004(pdf)
Barrier Integrity Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain MCPR within Operating Limits The team identified two examples of a Technical Specification 3.2.2, "Minimum Critical Power Ratio" (MCPR), non-cited violation for the failure to prevent transition boiling on the fuel during Operational Cycles 8 and 11. Fuel failures due to transition boiling were experienced during each cycle. Engineers failed to properly understand the affect of zinc injection on the cladding surfaces following the Cycle 8 fuel pin failures and zinc injection was reinitiated before the corrective actions to prevent recurrence were in place. The licensee had industry information that indicated that zinc injection contributed to the accumulation of loose crud and the formation of tenacious crud on the fuel. The additional crud rendered the Technical Specifications Minimum Critical Power Ratio (MCPR) calculations inaccurate and transition boiling occurred in localized areas. The licensee entered this issue into their corrective action program as CR-RBS-2006-0255.
The failure to prevent transition boiling in the core was a performance deficiency. The issue was more than minor because it impacted the barrier integrity cornerstone objective to maintain the integrity of the fuel cladding. The finding screened out as of very low safety significance (Green) because it only affected the fuel barrier. The issue had cross-cutting aspects in the areas of problem identification and resolution, in that the licensee failed to properly evaluate pertinent related industry information, which could have precluded the first violation, and failed to properly implement effective corrective measures in response to the first set of fuel failures, which led to the second violation.
Inspection Report# : 2005008(pdf)
Emergency Preparedness Significance: TBD May 10, 2006 Identified By: NRC Item Type: AV Apparent Violation Failure to Maintain a Standard Scheme of Emergency Classification and Action Levels in Use An apparent violation of 10 CFR 50.54(q) was identified for the licensees failure to ensure that adequate preplanned measures for Emergency Plan Emergency Action Levels were in place when seismic monitoring instrumentation was out of service at various times in 2004 and 2005. The seismic monitoring equipment was required to ensure the prompt implementation of the River Bend Emergency Plan as required by 10 CFR 50.54 (q) and the risk significant planning standard function,10 CFR 50.47(b)(4). The issue was entered into the licensees corrective action program as CR-RBS-2006-01283.
The finding was more than minor because it is associated with the procedure quality attribute of the Emergency Preparedness Cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Utilizing the Failure to Comply flow chart in Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, the inspectors determined that the finding was a failure to comply with an NRC requirement and was a Risk-Significant Planning Standard Problem involving a degraded Risk-Significant Planning Standard Function. The performance deficiency represents a degraded
 
2Q/2006 Inspection Findings - River Bend 1                                                                                                  Page 6 of 7 risk-significant planning standard function in that, during the periods that Reactor Mat Response Spectrum Recorder ERS-NBR2D or Free Field Seismic Trigger ERS-NBS4A were out of service, an existing Site Area Emergency emergency action level would not be declared. Based on the results of this evaluation, the finding was preliminarily determined to be of low to moderate safety significance.
Inspection Report# : 2006011(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for implementation of an EAL The NRC identified a noncited violation of 10 CFR Part 50, Appendix E, Section IV. B., as a result of inadequate procedures for the implementation of an emergency action level. The criteria in Procedure EIP-2-001, "Classification of Emergencies," Revision 12, for declaring an Alert emergency action level based on primary coolant leak rate, relied solely on a computer generated leakrate report that would not be valid under all conditions.
The licensee entered this finding into their corrective action program as CR-RBS-2005-03078 and issued Standing Order 192, as an interim corrective action, to provide additional criteria to determine whether a primary coolant leak rate Alert emergency action level declaration was required.
The finding is more than minor because it is associated with the Emergency Preparedness Cornerstone attribute of procedural quality and affects the cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inadequate procedure could result in a failure to declare an Alert emergency classification when required. Using Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," this finding was determined to be of very low safety significance since it was a failure to comply with a regulatory requirement associated with a risk-significant planning standard that did not result in the loss or degradation of that risk-significant planning standard function.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to control access to a high radiation area The inspector reviewed a self-revealing noncited violation of Technical Specification 5.7.1, resulting from the licensees failure to control access to a high radiation area. While transferring reverse osmosis system filters in the radwaste building, the licensee allowed two workers to inadvertently enter a high radiation area. This occurred after a guard prematurely left his post in front of the 123 foot elevation elevator door. The highest dose rate recorded by an electronic alarming dosimeter was 164 millirem per hour. The guard returned and evacuated the workers before they accrued additional radiation dose. Planned corrective action was still being evaluated by the licensee at the conclusion of the inspection.
The finding was more than minor because it was associated with the occupational radiation safety attribute of exposure control and affected the cornerstone objective in that not controlling a high radiation area could increase personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance in that the failure of the individual to guard the elevator door directly contributed to the violation.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform airborne radiation survey The inspector identified a noncited violation of 10 CFR 20.1501(a) because the licensee failed to survey airborne radioactivity. During the removal of local power range monitors, the licensee started collecting an air sample of the work area, but discarded the sample before analyzing it.
Successful passage through the portal monitors at the exit of the controlled access area confirmed that no worker experienced an uptake of radioactive material. Planned corrective action is still being evaluated.
The finding was more than minor because it was associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance in that the failure to maintain the sample for analysis directly contributed to the violation.
Inspection Report# : 2006003(pdf)
 
2Q/2006 Inspection Findings - River Bend 1                                                                                                Page 7 of 7 Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Jan 19, 2006 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Biannual Assessment The team reviewed approximately 225 condition reports, apparent and root cause analyses, as well as supporting documents to assess problem identification and resolution activities. In general, the corrective action program procedures and processes were effective, thresholds for identifying issues were low, and corrective actions were adequate to address conditions adverse to quality. Notwithstanding the above, poor engineering rigor associated with the prioritization and evaluation of issues resulted in a relatively high number of self-revealing and NRC identified findings. Some of these findings culminated in plant scrams and/or complicated operator response to emergency events. Others were related to equipment deficiencies, some of which resulted in inoperable safety-related equipment.
Based on the interviews conducted, the team concluded that a positive safety conscious work environment exists at River Bend Station. The team determined that employees felt free to raise safety concerns to station managers and supervisors, the employee concerns program, and the NRC.
However, the team received a few isolated comments regarding the correction action program feedback process. These individuals had previously identified corrective action issues and were not satisfied with the program's responses to their concerns. Some of these individuals commented that they were hesitant to use the corrective action program in the future. The licensee acknowledged the comments and planned to take action to address the concerns. All the interviewees believed that potential safety issues were being addressed.
Inspection Report# : 2005008(pdf)
Significance: N/A Sep 09, 2005 Identified By: NRC Item Type: FIN Finding Unplanned Scrams Exceed the Criteria for a White Performance Indicator The U.S. Nuclear Regulatory Commission performed this supplemental inspection to assess the licensee's evaluations associated with four unplanned reactor scrams that occurred between August 15, 2004 and January 15, 2005. The cumulative effect of these trips was that the Performance Indicator for unplanned scrams per 7000 critical hours crossed the threshold from Green (very low risk significance) to White (low to moderate risk significance) for the first quarter of calendar year 2005. The licensee performed individual root cause evaluations for all of the four reactor scrams. In addition to the individual trip evaluations, the licensee performed a common cause analysis to identify any performance and process issues that led to the White performance indicator. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that for each scram, the licensee performed a comprehensive and thorough evaluation in which specific problems were identified, an adequate root cause evaluation was performed, and corrective actions were taken or planned to prevent recurrence.
Inspection Report# : 2005012(pdf)
Last modified : August 25, 2006
 
3Q/2006 Inspection Findings - River Bend 1                                                                            Page 1 of 9 River Bend 1 3Q/2006 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to complete TS required actions within allowed completion time The NRC identified a noncited violation of Technical Specification 3.4.1.A for the licensee's failure to shut down one reactor recirculation loop within 2 hours of determining that jet pump loop flow mismatch was greater than 5 percent while operating at greater than 70 percent of rated core flow. On October 31, 2005, the Reactor Recirculation Flow Control Valve B hydraulic power unit tripped because of a blown control power fuse, causing Flow Control Valve B to drift open.
Operators throttled closed Flow Control Valve A to maintain reactor power at 100 percent, resulting in a jet pump loop flow mismatch of approximately 8.2 percent. The flow mismatch existed for 4.5 hours. The licensee entered this into their corrective action program as Condition Report CR-RBS-2006-00274.
The finding was more than minor because, if left uncorrected, it would become a more significant safety concern. Matched recirculation loop flows is an assumption used in the accident analysis for a loss of coolant accident resulting from a loop break. A flow mismatch could result in core response that is more severe than assumed in the accident analysis. The significance of this finding could not be evaluated using MC 0609, "Significance Determination Process." Based on management review, the finding was determined to be of very low safety significance based on the short duration of the flow mismatch, 4.5 hours, and the low likelihood of a loss of coolant accident during that time. The cause of this finding is related to the crosscutting element of human performance in that operators failed to implement Technical Specification requirements.
Inspection Report# : 2005005(pdf)
Mitigating Systems Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding Inadequate procedure for reassembly of the turbine bypass valve hydraulic system filter cartridge A self-revealing finding of very low safety significance was reviewed involving an inadequate procedure for conducting maintenance on the turbine bypass valve hydraulic system filter cartridge. This resulted in the improper reassembly of the filter. The resultant hydraulic oil leak caused the main turbine bypass valves to be inoperable, and a power reduction to less than 23.8 percent power was required by Technical Specifications. This issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2006-02632.
The performance deficiency associated with this finding was: (1) the failure to provide adequate instructions for reassembly of the turbine bypass valve hydraulic system filter cartridge to ensure that the cover gasket was properly installed, and (2) the failure to perform an adequate operational leak test of the system. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and it affected the associated cornerstone objective to ensure the availability and reliability of a system that responds to initiating events to prevent undesirable consequences. The inspectors performed a Phase 2 analysis using Manual Chapter 0609 and determined that the finding was of very low safety significance. The cause of the finding was related to the crosscutting element of human performance in that the licensee failed to provide complete, accurate, up-to-date instructions in the maintenance work package to change the hydraulic oil filter cartridge.
Inspection Report# : 2006004(pdf)
 
3Q/2006 Inspection Findings - River Bend 1                                                                          Page 2 of 9 Significance:      Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to identify Division III ESF bus supply breaker not racked in A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was reviewed involving the failure of the licensee to identify that the normal supply breaker to the Division III 4.16 kV engineered safety features bus was not properly racked in for a period of 24 days following maintenance. This issue was entered into the licensee's corrective action program as CR-RBS-2006-02402.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of configuration control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609, "Significance Determination Process," a Phase 3 analysis concluded that the finding was of very low safety significance.
The cause of the finding was related to the crosscutting aspect of problem identification and resolution in that the licensee failed to properly evaluate available indications to identify that the breaker was not properly racked in.
Inspection Report# : 2006003(pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately manage an increase in plant risk An NRC identified noncited violation of 10 CFR 50.65 Maintenance Rule Section (a)(4) was identified for the failure of the licensee to provide prescribed compensatory measures for two Orange shutdown risk conditions during Refueling Outage
: 13. Specifically, the preoutage risk assessment recommended that two work orders be in place for maintenance electricians to provide power to one spent fuel pool cooling pump in the event of problems with the running pump during periods of electrical bus maintenance. The inspectors found that the work packages were not in place before entering shutdown risk condition Orange on April 26, 2006, during the Division II engineering safety features bus testing, and May 3, 2006, during the Division I engineered safety features bus outage. This issue was entered into the licensee's corrective action program as CR-RBS-2006-01937.
The finding was more than minor because the licensee failed to implement a prescribed compensatory measure during the highest risk condition of Refueling Outage 13. The specific compensatory measures were called for in the preoutage risk assessment and the shutdown operations protection plan. The finding affected the mitigating system cornerstone because of the increased risk of a sustained loss of spent fuel pool cooling during core offloading operations. The finding could not be evaluated using the significance determination process, therefore the finding was reviewed by regional management and determined to be of very low safety significance. Factors that were considered included: (1) electrical maintenance technicians had previously performed the task of providing alternate power to a spent fuel pool cooling pump, (2) the necessary equipment was staged as part of the abnormal operating procedure for loss of decay heat removal, and (3) the relatively long time to boil of the spent fuel storage pool at that time during the refueling outage. The cause of the finding was related to the crosscutting aspect of human performance because the licensees planned maintenance activities and the predetermined increase in outage risk was not effectively managed by prescribed compensatory measures.
Inspection Report# : 2006003(pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure to verify required offsite power breaker alignment An NRC identified noncited violation of Technical Specification 5.4.1.a was identified for the failure of the licensee to provide an adequate surveillance test procedure to perform Technical Specification Surveillance Requirement 3.8.1.1.
Specifically, STP-000-0102, Power Distribution Alignment Check, Revision 4, did not verify the required offsite power circuit breaker alignment and indicated power availability for the Division III 4.16 kV engineered safety features bus as required in Modes 1, 2, and 3. This issue was entered into the licensee's corrective action program as CR-RBS-2006-02675 and -02402.
 
3Q/2006 Inspection Findings - River Bend 1                                                                          Page 3 of 9 The finding was more than minor because it was associated with the mitigating system cornerstone attribute of configuration control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609, "Significance Determination Process," a Phase 3 analysis concluded that the finding was of very low safety significance.
Inspection Report# : 2006003(pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Installation of Incorrect Relief Valve Caused Leak in Standby Service Water System A self-revealing noncited violation of Technical Specifications Section 5.4.1.a. was identified for the failure of procurement engineers to specify the correct replacement relief valve in a repetitive maintenance task to periodically replace thermal relief valves in the standby service water system. As a result, an incorrect valve was installed in the system which, following a system pressure transient, failed to reseat, creating a 10 gpm leak from the system. The valve was replaced and the issue was entered into the licensee's corrective action program as CR-RBS-2006-1054.
The finding is more than minor because it would become more significant if left uncorrected in that additional makeup to the standby service water system would be required during a sustained loss of off-site power. The finding affected the mitigating system cornerstone. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance because it did not result in the loss of the standby service water system safety function. The cause of the finding is related to the crosscutting element of problem identification and resolution because the problem which led to the installation of the incorrect valve had been previously identified and corrective actions were not effective in preventing recurrence.
Inspection Report# : 2006002(pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate maintenance results in a drywell steam leak from Low Pressure Coolant Injection Train A Testable Check Valve A self-revealing noncited violation of Technical Specification Section 5.4.1.a, was identified for the failure of mechanical maintenance technicians to correctly reassemble Low Pressure Coolant Injection Testable Check Valve E12-AOVF041A during Refueling Outage 12. As a result, a steam leak from a valve flange caused a rise in drywell unidentified leakage. The issue was entered into the licensees corrective action program as CR-RBS-2006-00546 and the valve was repaired.
The finding is more than minor because it would have become a more significant safety concern if left uncorrected. The leakage would have continued to increase during the cycle, and it would have continued to have an adverse affect on indicated reactor vessel level. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance because it did not result in a loss of the low pressure coolant injection system safety function and was not potentially risk significant due to seismic, flooding, or severe weather related initiating events. The finding had crosscutting aspects associated with human performance in that maintenance technicians incorrectly reassembled the valve during refueling outage 12.
Inspection Report# : 2006002(pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadvertent Initiation of High Pressure Core Spray Caused by the Use of the Wrong Test Plug During Surveillance Testing A self-revealing noncited violation of Technical Specifications Section 5.4.1.a. was identified for the failure to provide adequate procedural guidance for the use of a test plug during the performance of a required surveillance test procedure.
The use of the wrong test plug caused an initiation of the high pressure core spray system and injection into the vessel. The issue was entered in the licensees corrective action program as CR-RBS-2006-00283.
 
3Q/2006 Inspection Findings - River Bend 1                                                                              Page 4 of 9 The finding is more than minor because it is associated with the mitigating system cornerstone attribute of equipment performance and the cornerstone objective to ensure the availability and reliability of high pressure core spray, a system that responds to initiating events to prevent undesirable consequences. The Phase 1 worksheets in Manual Chapter 0609, "Significance Determination Process," were used to conclude that a Phase 2 analysis was required because there was an actual loss of system safety function. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding is related to the crosscutting element of human performance because the technicians did not verify that they were using the correct test plug for the surveillance test being performed.
Inspection Report# : 2006002(pdf)
Significance:      Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable Preconditioning of a Safety-related Valve Prior to Surveillance Testing The team identified a non-cited violation of Technical Specification 5.4.1.a (Procedures) for unacceptable preconditioning of a low pressure core spray keepfill system check valve. The test procedure failed to prescribe testing the check valve in the as-found condition. Instead (during testing of the system pump) the document directed operators to flush the valve at 27 gpm for up to 20 minutes prior to the check valve test. Corrosion buildup in the valve, which had previously caused valve failures, was a known concern and the preconditioning could have masked performance problems. Failure of the valve to perform its safety function puts the low pressure core spray system at risk of water hammer during a loss of offsite power event. The licensee planned to test the valve in the as-found configuration during future tests. The licensee documented this issue in their corrective action program as CR-RBS-2005-04123.
The failure to properly test the subject check valve was a performance deficiency. The finding was more than minor because, if left uncorrected, the problem could result in a more significant safety concern. Specifically, the surveillance test may not identify valve failure. The finding was of very low risk significance because it was not a design/qualification issue, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its technical specification allowable outage time, did not result in a loss of function of non safety-related risk significant equipment and was not risk significant due to external events. The finding had problem identification and resolution cross-cutting aspects because the licensee had failed to properly evaluate the issue as preconditioning in response to readily available industry information.
Inspection Report# : 2005008(pdf)
Significance:      Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Replacement of a Valve to Correct a Significant Condition Adverse to Quality The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Actions) for the failure to take prompt corrective measures to address a significant condition adverse to quality. Specifically, the low pressure core spray keepfill pump discharge check valve failed on two occasions (significant conditions adverse to quality) and planned corrective measures to replace the check valve were not timely. The check valve failures put the low pressure core spray system at increased water hammer risk during a loss of offsite power event. The licensee had identified that corrosion buildup was causing the valve to leak excessively when closed. The licensee documented this issue in their corrective action program as CR-RBS-2005-04162 and planned to replace the valve at the next available opportunity.
The failure to take prompt corrective measures to address a significant condition adverse to quality was a performance deficiency. The finding was greater than minor because it was an equipment performance reliability issue which impacted the mitigating systems cornerstone objective to ensure the reliability of systems that respond to initiating events. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the finding was of very low risk significance because it was not a design/qualification issue, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its technical specification allowable outage time, did not result in a loss of function of non safety-related risk significant equipment and was not risk significant due to external events. The finding had cross-cutting aspects in the area of problem identification and resolution.
Inspection Report# : 2005008(pdf)
 
3Q/2006 Inspection Findings - River Bend 1                                                                              Page 5 of 9 Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Set MOV Limit Switches in Accordance with Design Documentation The team identified a 10 CFR 50, Appendix B, Criterion V (procedures) non-cited violation for the failure to set safety-related limit switches in accordance documents appropriate to the circumstances for 34 safety-related throttle valves. The licensee set motor-operated valve (MOV) open indication light limit switches so that the open indication de-energized between the 95% and 100% closed positions, whereas the applicable procedure and design drawing required that the limit switches be set to the 100% closed position. This practice had caused repetitive operational problems in the plant. The licensee entered this issue into their corrective action program as CR-RBS-2005-04113.
The failure to adjust MOV limit switches in accordance with documents appropriate to the circumstances was a performance deficiency. The issue was more than minor because it affected the mitigating systems cornerstone objective, in that it affected the operability, availability, reliability or function of a system or train in a mitigating system. The finding was of very low safety significance because it was a design/qualification deficiency confirmed not to result in loss of operability per "Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment." This finding had cross-cutting aspects in the areas of human performance, (the failure to follow procedures) and problem identification and resolution because the licensee failed to identify the problem in response to a prior related NRC violation.
Inspection Report# : 2005008(pdf)
Significance:        Dec 31, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate design assumption results in RCIC turbine exhaust header filling with water following an automatic high water level shutdown A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensee's failure to address the worst case conditions in the sizing calculation for the reactor core isolation cooling turbine exhaust line vacuum breaker system as part of a plant modification to remove the internals of the reactor core isolation cooling turbine exhaust line check valve. As a result, on December 10, 2004, when the reactor core isolation cooling system was started and subsequently shutdown on high reactor water level following a scram and loss of feedwater, the turbine exhaust line filled with water from the suppression pool, causing the operators to consider the system unavailable and complicating their response to the event. The licensee entered this finding into their corrective action program as CR-RBS-2005-00724 and reinstalled the turbine exhaust line check valve internals in February 2005.
The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective to ensure the availability and reliability of the reactor core isolation cooling system, a system that responds to initiating events (loss of feedwater and station blackout), to prevent undesirable consequences. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because it represented a design deficiency that did not result in a loss of system function.
Inspection Report# : 2005005(pdf)
Barrier Integrity Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain MCPR within Operating Limits The team identified two examples of a Technical Specification 3.2.2, "Minimum Critical Power Ratio" (MCPR), non-cited violation for the failure to prevent transition boiling on the fuel during Operational Cycles 8 and 11. Fuel failures due to
 
3Q/2006 Inspection Findings - River Bend 1                                                                            Page 6 of 9 transition boiling were experienced during each cycle. Engineers failed to properly understand the affect of zinc injection on the cladding surfaces following the Cycle 8 fuel pin failures and zinc injection was reinitiated before the corrective actions to prevent recurrence were in place. The licensee had industry information that indicated that zinc injection contributed to the accumulation of loose crud and the formation of tenacious crud on the fuel. The additional crud rendered the Technical Specifications Minimum Critical Power Ratio (MCPR) calculations inaccurate and transition boiling occurred in localized areas. The licensee entered this issue into their corrective action program as CR-RBS-2006-0255.
The failure to prevent transition boiling in the core was a performance deficiency. The issue was more than minor because it impacted the barrier integrity cornerstone objective to maintain the integrity of the fuel cladding. The finding screened out as of very low safety significance (Green) because it only affected the fuel barrier. The issue had cross-cutting aspects in the areas of problem identification and resolution, in that the licensee failed to properly evaluate pertinent related industry information, which could have precluded the first violation, and failed to properly implement effective corrective measures in response to the first set of fuel failures, which led to the second violation.
Inspection Report# : 2005008(pdf)
Emergency Preparedness Significance:      May 10, 2006 Identified By: NRC Item Type: AV Apparent Violation Failure to Maintain a Standard Scheme of Emergency Classification and Action Levels in Use The NRC has considered the information developed during the inspection, the EOI position on the issue which was attached to the inspection report, the information you provided at the Regulatory Conference, and the information provided by your staff in a {{letter dated|date=July 28, 2006|text=July 28, 2006, letter}} following the conference. On the basis of this information, the NRC has concluded that a violation occurred. The violation involves a failure to meet 10 CFR 50.54(q), which requires that the licensee follow and maintain in effect emergency plans which meet the standards in 10 CFR 50.47(b). Specifically, during periods when certain seismic monitoring instrumentation was out of service, the licensee could not implement the emergency actions levels as described in the applicable Emergency Plan implementing procedure. However, there was other seismic instrumentation available during these periods that could be used to determine the ground force acceleration associated with a seismic event in the vicinity of the River Bend Station. This information could then be used by the Operations Shift Manager or Emergency Director to determine the correct classification for a seismic event; although, the classification could be delayed for as long as 4 hours. On the basis of this information, the NRC has concluded that the inspection finding did not represent a degradation of a risk significant planning standard function, as defined in Appendix B of NRCs Inspection Manual Chapter 0609, and therefore is of very low safety significance.
Also, this finding had crosscutting aspects in the area of problem identification and resolution because the River Bend Station staff did not identify the effect that inoperable seismic monitoring instrumentation had on the ability to implement the River Bend Station Emergency Plan and did not effectively utilize pertinent industry operating experience to prevent the performance deficiency.
Inspection Report# : 2006011(pdf)
Significance:      Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for implementation of an EAL The NRC identified a noncited violation of 10 CFR Part 50, Appendix E, Section IV. B., as a result of inadequate procedures for the implementation of an emergency action level. The criteria in Procedure EIP-2-001, "Classification of Emergencies," Revision 12, for declaring an Alert emergency action level based on primary coolant leak rate, relied solely on a computer generated leakrate report that would not be valid under all conditions. The licensee entered this finding into their corrective action program as CR-RBS-2005-03078 and issued Standing Order 192, as an interim corrective action, to provide additional criteria to determine whether a primary coolant leak rate Alert emergency action level declaration was required.
 
3Q/2006 Inspection Findings - River Bend 1                                                                            Page 7 of 9 The finding is more than minor because it is associated with the Emergency Preparedness Cornerstone attribute of procedural quality and affects the cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inadequate procedure could result in a failure to declare an Alert emergency classification when required. Using Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," this finding was determined to be of very low safety significance since it was a failure to comply with a regulatory requirement associated with a risk-significant planning standard that did not result in the loss or degradation of that risk-significant planning standard function.
Inspection Report# : 2005005(pdf)
Occupational Radiation Safety Significance:      Sep 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow radiation work permit requirements The inspector reviewed a self-revealing, noncited violation of Technical Specification 5.4.1 resulting from workers' failure to follow radiation work permit requirements. Two workers performing a scaffolding modification in the inclined fuel transfer system canal became externally and internally contaminated. As the workers were exiting the controlled access area, they alarmed the personnel contamination monitors. Based upon the whole-body count results, the licensee assigned a committed effective dose equivalent of 30 millirem to one worker and 70 millirem to the other worker. The licensee's investigation determined that the workers did not inform radiation protection personnel that they would be lowering the scaffolding 3 feet below surveyed areas and contamination control devices. Consequently, the workers were in radiological conditions not bounded by the radiation work permit and as low as is reasonably achievable planners did not have a chance to conduct a total effective dose equivalent as low as is reasonably achievable review to determine if respiratory protection was necessary. As a corrective action, the licensee is incorporating a lessons learned item associated with this event into radiation worker training.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure/contamination control) and affects the Occupational Radiation Safety cornerstone objective in that the failure to follow radiation work permit instructions resulted in additional personnel exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because the workers failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2006004(pdf)
Significance:      Sep 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to evaluate radiological conditions The inspector reviewed a self-revealing, noncited violation of 10 CFR 20.1501(a) resulting from the licensee's failure to correctly measure the airborne radioactivity where personnel worked. The licensee's review of the January 26, 2006, contamination event identified that the air sample taken to support the work activity was positioned above the high-efficiency particulate air hose suction in an air flow area above the actual work area. This meant that the air sample was not representative of the workers' actual work area. In addition, the radiation protection technician providing continuous job coverage failed to identify the deficiency and adjust the position of the air sampler. As a corrective action, the licensee is incorporating a lessons learned item associated with this event into the radiation protection technician training.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affects the Occupational Radiation Safety cornerstone objective in that an inadequate evaluation of the hazards could lead to inadequate radiation protection and dose saving measures. This finding could also be reasonably viewed as a precursor to
 
3Q/2006 Inspection Findings - River Bend 1                                                                          Page 8 of 9 a significant event, such as a personnel overexposure, had contamination levels been higher. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because radiation protection personnel failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2006004(pdf)
Significance:        Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to evaluate radiological conditions The inspector identified a noncited violation of 10 CFR 20.1501(a), resulting from the licensees use of an inadequate alpha contamination survey technique. The inspector determined that the licensee's procedure for the use of the Eberline SAC-4 alpha scintillation counter established a screening limit that did not allow sufficient sample activity for the discovery of alpha emitting radionuclides. Therefore, the inspector concluded that surveys using this technique could not identify alpha contamination and were inadequate. As a corrective action, the licensee adapted the corporate procedural guidance, which raised the maximum sample activity.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affects the Occupational Radiation Safety cornerstone objective in that an inadequate evaluation of the hazards could lead to inadequate radiation protection and dose saving measures. This finding could also be reasonably viewed as a precursor to a significant event, such as a personnel overexposure, had contamination levels been higher. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because the licensee used procedures that were inadequate to ensure that alpha contamination was identified.
Inspection Report# : 2006004(pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to control access to a high radiation area The inspector reviewed a self-revealing noncited violation of Technical Specification 5.7.1, resulting from the licensees failure to control access to a high radiation area. While transferring reverse osmosis system filters in the radwaste building, the licensee allowed two workers to inadvertently enter a high radiation area. This occurred after a guard prematurely left his post in front of the 123 foot elevation elevator door. The highest dose rate recorded by an electronic alarming dosimeter was 164 millirem per hour. The guard returned and evacuated the workers before they accrued additional radiation dose.
Planned corrective action was still being evaluated by the licensee at the conclusion of the inspection.
The finding was more than minor because it was associated with the occupational radiation safety attribute of exposure control and affected the cornerstone objective in that not controlling a high radiation area could increase personal exposure.
Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance in that the failure of the individual to guard the elevator door directly contributed to the violation.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform airborne radiation survey
 
3Q/2006 Inspection Findings - River Bend 1                                                                          Page 9 of 9 The inspector identified a noncited violation of 10 CFR 20.1501(a) because the licensee failed to survey airborne radioactivity. During the removal of local power range monitors, the licensee started collecting an air sample of the work area, but discarded the sample before analyzing it. Successful passage through the portal monitors at the exit of the controlled access area confirmed that no worker experienced an uptake of radioactive material. Planned corrective action is still being evaluated.
The finding was more than minor because it was associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance in that the failure to maintain the sample for analysis directly contributed to the violation.
Inspection Report# : 2006003(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Jan 19, 2006 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Biannual Assessment The team reviewed approximately 225 condition reports, apparent and root cause analyses, as well as supporting documents to assess problem identification and resolution activities. In general, the corrective action program procedures and processes were effective, thresholds for identifying issues were low, and corrective actions were adequate to address conditions adverse to quality. Notwithstanding the above, poor engineering rigor associated with the prioritization and evaluation of issues resulted in a relatively high number of self-revealing and NRC identified findings. Some of these findings culminated in plant scrams and/or complicated operator response to emergency events. Others were related to equipment deficiencies, some of which resulted in inoperable safety-related equipment.
Based on the interviews conducted, the team concluded that a positive safety conscious work environment exists at River Bend Station. The team determined that employees felt free to raise safety concerns to station managers and supervisors, the employee concerns program, and the NRC. However, the team received a few isolated comments regarding the correction action program feedback process. These individuals had previously identified corrective action issues and were not satisfied with the program's responses to their concerns. Some of these individuals commented that they were hesitant to use the corrective action program in the future. The licensee acknowledged the comments and planned to take action to address the concerns. All the interviewees believed that potential safety issues were being addressed.
Inspection Report# : 2005008(pdf)
Last modified : December 21, 2006
 
4Q/2006 Inspection Findings - River Bend 1                                                                            Page 1 of 8 River Bend 1 4Q/2006 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding Inadequate procedure for reassembly of the turbine bypass valve hydraulic system filter cartridge A self-revealing finding of very low safety significance was reviewed involving an inadequate procedure for conducting maintenance on the turbine bypass valve hydraulic system filter cartridge. This resulted in the improper reassembly of the filter. The resultant hydraulic oil leak caused the main turbine bypass valves to be inoperable, and a power reduction to less than 23.8 percent power was required by Technical Specifications. This issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2006-02632.
The performance deficiency associated with this finding was: (1) the failure to provide adequate instructions for reassembly of the turbine bypass valve hydraulic system filter cartridge to ensure that the cover gasket was properly installed, and (2) the failure to perform an adequate operational leak test of the system. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and it affected the associated cornerstone objective to ensure the availability and reliability of a system that responds to initiating events to prevent undesirable consequences. The inspectors performed a Phase 2 analysis using Manual Chapter 0609 and determined that the finding was of very low safety significance. The cause of the finding was related to the crosscutting element of human performance in that the licensee failed to provide complete, accurate, up-to-date instructions in the maintenance work package to change the hydraulic oil filter cartridge.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to identify Division III ESF bus supply breaker not racked in A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was reviewed involving the failure of the licensee to identify that the normal supply breaker to the Division III 4.16 kV engineered safety features bus was not properly racked in for a period of 24 days following maintenance. This issue was entered into the licensee's corrective action program as CR-RBS-2006-02402.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of configuration control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609, "Significance Determination Process," a Phase 3 analysis concluded that the finding was of very low safety significance.
The cause of the finding was related to the crosscutting aspect of problem identification and resolution in that the licensee failed to properly evaluate available indications to identify that the breaker was not properly racked in.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2006 Inspection Findings - River Bend 1                                                                          Page 2 of 8 Failure to adequately manage an increase in plant risk An NRC identified noncited violation of 10 CFR 50.65 Maintenance Rule Section (a)(4) was identified for the failure of the licensee to provide prescribed compensatory measures for two Orange shutdown risk conditions during Refueling Outage
: 13. Specifically, the preoutage risk assessment recommended that two work orders be in place for maintenance electricians to provide power to one spent fuel pool cooling pump in the event of problems with the running pump during periods of electrical bus maintenance. The inspectors found that the work packages were not in place before entering shutdown risk condition Orange on April 26, 2006, during the Division II engineering safety features bus testing, and May 3, 2006, during the Division I engineered safety features bus outage. This issue was entered into the licensee's corrective action program as CR-RBS-2006-01937.
The finding was more than minor because the licensee failed to implement a prescribed compensatory measure during the highest risk condition of Refueling Outage 13. The specific compensatory measures were called for in the preoutage risk assessment and the shutdown operations protection plan. The finding affected the mitigating system cornerstone because of the increased risk of a sustained loss of spent fuel pool cooling during core offloading operations. The finding could not be evaluated using the significance determination process, therefore the finding was reviewed by regional management and determined to be of very low safety significance. Factors that were considered included: (1) electrical maintenance technicians had previously performed the task of providing alternate power to a spent fuel pool cooling pump, (2) the necessary equipment was staged as part of the abnormal operating procedure for loss of decay heat removal, and (3) the relatively long time to boil of the spent fuel storage pool at that time during the refueling outage. The cause of the finding was related to the crosscutting aspect of human performance because the licensees planned maintenance activities and the predetermined increase in outage risk was not effectively managed by prescribed compensatory measures.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure to verify required offsite power breaker alignment An NRC identified noncited violation of Technical Specification 5.4.1.a was identified for the failure of the licensee to provide an adequate surveillance test procedure to perform Technical Specification Surveillance Requirement 3.8.1.1.
Specifically, STP-000-0102, Power Distribution Alignment Check, Revision 4, did not verify the required offsite power circuit breaker alignment and indicated power availability for the Division III 4.16 kV engineered safety features bus as required in Modes 1, 2, and 3. This issue was entered into the licensee's corrective action program as CR-RBS-2006-02675 and -02402.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of configuration control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609, "Significance Determination Process," a Phase 3 analysis concluded that the finding was of very low safety significance.
Inspection Report# : 2006003 (pdf)
Significance:      Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Installation of Incorrect Relief Valve Caused Leak in Standby Service Water System A self-revealing noncited violation of Technical Specifications Section 5.4.1.a. was identified for the failure of procurement engineers to specify the correct replacement relief valve in a repetitive maintenance task to periodically replace thermal relief valves in the standby service water system. As a result, an incorrect valve was installed in the system which, following a system pressure transient, failed to reseat, creating a 10 gpm leak from the system. The valve was replaced and the issue was entered into the licensee's corrective action program as CR-RBS-2006-1054.
The finding is more than minor because it would become more significant if left uncorrected in that additional makeup to the standby service water system would be required during a sustained loss of off-site power. The finding affected the mitigating system cornerstone. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance because it did not result in the loss of the standby service water system safety function. The cause of the finding is related to the crosscutting element of problem identification and
 
4Q/2006 Inspection Findings - River Bend 1                                                                          Page 3 of 8 resolution because the problem which led to the installation of the incorrect valve had been previously identified and corrective actions were not effective in preventing recurrence.
Inspection Report# : 2006002 (pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate maintenance results in a drywell steam leak from Low Pressure Coolant Injection Train A Testable Check Valve A self-revealing noncited violation of Technical Specification Section 5.4.1.a, was identified for the failure of mechanical maintenance technicians to correctly reassemble Low Pressure Coolant Injection Testable Check Valve E12-AOVF041A during Refueling Outage 12. As a result, a steam leak from a valve flange caused a rise in drywell unidentified leakage. The issue was entered into the licensees corrective action program as CR-RBS-2006-00546 and the valve was repaired.
The finding is more than minor because it would have become a more significant safety concern if left uncorrected. The leakage would have continued to increase during the cycle, and it would have continued to have an adverse affect on indicated reactor vessel level. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance because it did not result in a loss of the low pressure coolant injection system safety function and was not potentially risk significant due to seismic, flooding, or severe weather related initiating events. The finding had crosscutting aspects associated with human performance in that maintenance technicians incorrectly reassembled the valve during refueling outage 12.
Inspection Report# : 2006002 (pdf)
Significance:        Mar 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadvertent Initiation of High Pressure Core Spray Caused by the Use of the Wrong Test Plug During Surveillance Testing A self-revealing noncited violation of Technical Specifications Section 5.4.1.a. was identified for the failure to provide adequate procedural guidance for the use of a test plug during the performance of a required surveillance test procedure.
The use of the wrong test plug caused an initiation of the high pressure core spray system and injection into the vessel. The issue was entered in the licensees corrective action program as CR-RBS-2006-00283.
The finding is more than minor because it is associated with the mitigating system cornerstone attribute of equipment performance and the cornerstone objective to ensure the availability and reliability of high pressure core spray, a system that responds to initiating events to prevent undesirable consequences. The Phase 1 worksheets in Manual Chapter 0609, "Significance Determination Process," were used to conclude that a Phase 2 analysis was required because there was an actual loss of system safety function. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding is related to the crosscutting element of human performance because the technicians did not verify that they were using the correct test plug for the surveillance test being performed.
Inspection Report# : 2006002 (pdf)
Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable Preconditioning of a Safety-related Valve Prior to Surveillance Testing The team identified a non-cited violation of Technical Specification 5.4.1.a (Procedures) for unacceptable preconditioning of a low pressure core spray keepfill system check valve. The test procedure failed to prescribe testing the check valve in the as-found condition. Instead (during testing of the system pump) the document directed operators to flush the valve at 27 gpm for up to 20 minutes prior to the check valve test. Corrosion buildup in the valve, which had previously caused valve failures, was a known concern and the preconditioning could have masked performance problems. Failure of the valve to perform its safety function puts the low pressure core spray system at risk of water hammer during a loss of offsite power event. The licensee planned to test the valve in the as-found configuration during future tests. The licensee documented this issue in their corrective action program as CR-RBS-2005-04123.
 
4Q/2006 Inspection Findings - River Bend 1                                                                              Page 4 of 8 The failure to properly test the subject check valve was a performance deficiency. The finding was more than minor because, if left uncorrected, the problem could result in a more significant safety concern. Specifically, the surveillance test may not identify valve failure. The finding was of very low risk significance because it was not a design/qualification issue, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its technical specification allowable outage time, did not result in a loss of function of non safety-related risk significant equipment and was not risk significant due to external events. The finding had problem identification and resolution cross-cutting aspects because the licensee had failed to properly evaluate the issue as preconditioning in response to readily available industry information.
Inspection Report# : 2005008 (pdf)
Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Replacement of a Valve to Correct a Significant Condition Adverse to Quality The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Actions) for the failure to take prompt corrective measures to address a significant condition adverse to quality. Specifically, the low pressure core spray keepfill pump discharge check valve failed on two occasions (significant conditions adverse to quality) and planned corrective measures to replace the check valve were not timely. The check valve failures put the low pressure core spray system at increased water hammer risk during a loss of offsite power event. The licensee had identified that corrosion buildup was causing the valve to leak excessively when closed. The licensee documented this issue in their corrective action program as CR-RBS-2005-04162 and planned to replace the valve at the next available opportunity.
The failure to take prompt corrective measures to address a significant condition adverse to quality was a performance deficiency. The finding was greater than minor because it was an equipment performance reliability issue which impacted the mitigating systems cornerstone objective to ensure the reliability of systems that respond to initiating events. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the finding was of very low risk significance because it was not a design/qualification issue, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its technical specification allowable outage time, did not result in a loss of function of non safety-related risk significant equipment and was not risk significant due to external events. The finding had cross-cutting aspects in the area of problem identification and resolution.
Inspection Report# : 2005008 (pdf)
Significance:        Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Set MOV Limit Switches in Accordance with Design Documentation The team identified a 10 CFR 50, Appendix B, Criterion V (procedures) non-cited violation for the failure to set safety-related limit switches in accordance documents appropriate to the circumstances for 34 safety-related throttle valves. The licensee set motor-operated valve (MOV) open indication light limit switches so that the open indication de-energized between the 95% and 100% closed positions, whereas the applicable procedure and design drawing required that the limit switches be set to the 100% closed position. This practice had caused repetitive operational problems in the plant. The licensee entered this issue into their corrective action program as CR-RBS-2005-04113.
The failure to adjust MOV limit switches in accordance with documents appropriate to the circumstances was a performance deficiency. The issue was more than minor because it affected the mitigating systems cornerstone objective, in that it affected the operability, availability, reliability or function of a system or train in a mitigating system. The finding was of very low safety significance because it was a design/qualification deficiency confirmed not to result in loss of operability per "Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment." This finding had cross-cutting aspects in the areas of human performance, (the failure to follow procedures) and problem identification and resolution because the licensee failed to identify the problem in response to a prior related NRC violation.
Inspection Report# : 2005008 (pdf)
 
4Q/2006 Inspection Findings - River Bend 1                                                                            Page 5 of 8 Barrier Integrity Significance:      Jan 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain MCPR within Operating Limits The team identified two examples of a Technical Specification 3.2.2, "Minimum Critical Power Ratio" (MCPR), non-cited violation for the failure to prevent transition boiling on the fuel during Operational Cycles 8 and 11. Fuel failures due to transition boiling were experienced during each cycle. Engineers failed to properly understand the affect of zinc injection on the cladding surfaces following the Cycle 8 fuel pin failures and zinc injection was reinitiated before the corrective actions to prevent recurrence were in place. The licensee had industry information that indicated that zinc injection contributed to the accumulation of loose crud and the formation of tenacious crud on the fuel. The additional crud rendered the Technical Specifications Minimum Critical Power Ratio (MCPR) calculations inaccurate and transition boiling occurred in localized areas. The licensee entered this issue into their corrective action program as CR-RBS-2006-0255.
The failure to prevent transition boiling in the core was a performance deficiency. The issue was more than minor because it impacted the barrier integrity cornerstone objective to maintain the integrity of the fuel cladding. The finding screened out as of very low safety significance (Green) because it only affected the fuel barrier. The issue had cross-cutting aspects in the areas of problem identification and resolution, in that the licensee failed to properly evaluate pertinent related industry information, which could have precluded the first violation, and failed to properly implement effective corrective measures in response to the first set of fuel failures, which led to the second violation.
Inspection Report# : 2005008 (pdf)
Emergency Preparedness Significance:      May 10, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain a Standard Scheme of Emergency Classification and Action Levels in Use The NRC has considered the information developed during the inspection, the EOI position on the issue which was attached to the inspection report, the information you provided at the Regulatory Conference, and the information provided by your staff in a {{letter dated|date=July 28, 2006|text=July 28, 2006, letter}} following the conference. On the basis of this information, the NRC has concluded that a violation occurred. The violation involves a failure to meet 10 CFR 50.54(q), which requires that the licensee follow and maintain in effect emergency plans which meet the standards in 10 CFR 50.47(b). Specifically, during periods when certain seismic monitoring instrumentation was out of service, the licensee could not implement the emergency actions levels as described in the applicable Emergency Plan implementing procedure. However, there was other seismic instrumentation available during these periods that could be used to determine the ground force acceleration associated with a seismic event in the vicinity of the River Bend Station. This information could then be used by the Operations Shift Manager or Emergency Director to determine the correct classification for a seismic event; although, the classification could be delayed for as long as 4 hours. On the basis of this information, the NRC has concluded that the inspection finding did not represent a degradation of a risk significant planning standard function, as defined in Appendix B of NRCs Inspection Manual Chapter 0609, and therefore is of very low safety significance.
Also, this finding had crosscutting aspects in the area of problem identification and resolution because the River Bend Station staff did not identify the effect that inoperable seismic monitoring instrumentation had on the ability to implement the River Bend Station Emergency Plan and did not effectively utilize pertinent industry operating experience to prevent the performance deficiency.
Inspection Report# : 2006005 (pdf)
Inspection Report# : 2006011 (pdf)
 
4Q/2006 Inspection Findings - River Bend 1                                                                            Page 6 of 8 Occupational Radiation Safety Significance:      Sep 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow radiation work permit requirements The inspector reviewed a self-revealing, noncited violation of Technical Specification 5.4.1 resulting from workers' failure to follow radiation work permit requirements. Two workers performing a scaffolding modification in the inclined fuel transfer system canal became externally and internally contaminated. As the workers were exiting the controlled access area, they alarmed the personnel contamination monitors. Based upon the whole-body count results, the licensee assigned a committed effective dose equivalent of 30 millirem to one worker and 70 millirem to the other worker. The licensee's investigation determined that the workers did not inform radiation protection personnel that they would be lowering the scaffolding 3 feet below surveyed areas and contamination control devices. Consequently, the workers were in radiological conditions not bounded by the radiation work permit and as low as is reasonably achievable planners did not have a chance to conduct a total effective dose equivalent as low as is reasonably achievable review to determine if respiratory protection was necessary. As a corrective action, the licensee is incorporating a lessons learned item associated with this event into radiation worker training.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure/contamination control) and affects the Occupational Radiation Safety cornerstone objective in that the failure to follow radiation work permit instructions resulted in additional personnel exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because the workers failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2006004 (pdf)
Significance:      Sep 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to evaluate radiological conditions The inspector reviewed a self-revealing, noncited violation of 10 CFR 20.1501(a) resulting from the licensee's failure to correctly measure the airborne radioactivity where personnel worked. The licensee's review of the January 26, 2006, contamination event identified that the air sample taken to support the work activity was positioned above the high-efficiency particulate air hose suction in an air flow area above the actual work area. This meant that the air sample was not representative of the workers' actual work area. In addition, the radiation protection technician providing continuous job coverage failed to identify the deficiency and adjust the position of the air sampler. As a corrective action, the licensee is incorporating a lessons learned item associated with this event into the radiation protection technician training.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affects the Occupational Radiation Safety cornerstone objective in that an inadequate evaluation of the hazards could lead to inadequate radiation protection and dose saving measures. This finding could also be reasonably viewed as a precursor to a significant event, such as a personnel overexposure, had contamination levels been higher. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because radiation protection personnel failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2006004 (pdf)
Significance:      Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2006 Inspection Findings - River Bend 1                                                                          Page 7 of 8 Failure to evaluate radiological conditions The inspector identified a noncited violation of 10 CFR 20.1501(a), resulting from the licensees use of an inadequate alpha contamination survey technique. The inspector determined that the licensee's procedure for the use of the Eberline SAC-4 alpha scintillation counter established a screening limit that did not allow sufficient sample activity for the discovery of alpha emitting radionuclides. Therefore, the inspector concluded that surveys using this technique could not identify alpha contamination and were inadequate. As a corrective action, the licensee adapted the corporate procedural guidance, which raised the maximum sample activity.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affects the Occupational Radiation Safety cornerstone objective in that an inadequate evaluation of the hazards could lead to inadequate radiation protection and dose saving measures. This finding could also be reasonably viewed as a precursor to a significant event, such as a personnel overexposure, had contamination levels been higher. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because the licensee used procedures that were inadequate to ensure that alpha contamination was identified.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to control access to a high radiation area The inspector reviewed a self-revealing noncited violation of Technical Specification 5.7.1, resulting from the licensees failure to control access to a high radiation area. While transferring reverse osmosis system filters in the radwaste building, the licensee allowed two workers to inadvertently enter a high radiation area. This occurred after a guard prematurely left his post in front of the 123 foot elevation elevator door. The highest dose rate recorded by an electronic alarming dosimeter was 164 millirem per hour. The guard returned and evacuated the workers before they accrued additional radiation dose.
Planned corrective action was still being evaluated by the licensee at the conclusion of the inspection.
The finding was more than minor because it was associated with the occupational radiation safety attribute of exposure control and affected the cornerstone objective in that not controlling a high radiation area could increase personal exposure.
Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance in that the failure of the individual to guard the elevator door directly contributed to the violation.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform airborne radiation survey The inspector identified a noncited violation of 10 CFR 20.1501(a) because the licensee failed to survey airborne radioactivity. During the removal of local power range monitors, the licensee started collecting an air sample of the work area, but discarded the sample before analyzing it. Successful passage through the portal monitors at the exit of the controlled access area confirmed that no worker experienced an uptake of radioactive material. Planned corrective action is still being evaluated.
The finding was more than minor because it was associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance in that the failure to maintain
 
4Q/2006 Inspection Findings - River Bend 1                                                                        Page 8 of 8 the sample for analysis directly contributed to the violation.
Inspection Report# : 2006003 (pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Significance: N/A Jan 19, 2006 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution Biannual Assessment The team reviewed approximately 225 condition reports, apparent and root cause analyses, as well as supporting documents to assess problem identification and resolution activities. In general, the corrective action program procedures and processes were effective, thresholds for identifying issues were low, and corrective actions were adequate to address conditions adverse to quality. Notwithstanding the above, poor engineering rigor associated with the prioritization and evaluation of issues resulted in a relatively high number of self-revealing and NRC identified findings. Some of these findings culminated in plant scrams and/or complicated operator response to emergency events. Others were related to equipment deficiencies, some of which resulted in inoperable safety-related equipment.
Based on the interviews conducted, the team concluded that a positive safety conscious work environment exists at River Bend Station. The team determined that employees felt free to raise safety concerns to station managers and supervisors, the employee concerns program, and the NRC. However, the team received a few isolated comments regarding the correction action program feedback process. These individuals had previously identified corrective action issues and were not satisfied with the program's responses to their concerns. Some of these individuals commented that they were hesitant to use the corrective action program in the future. The licensee acknowledged the comments and planned to take action to address the concerns. All the interviewees believed that potential safety issues were being addressed.
Inspection Report# : 2005008 (pdf)
Last modified : March 01, 2007
 
River Bend 1 1Q/2007 Plant Inspection Findings Initiating Events Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct discrepancies between the design function and observed response of the feedwater isolation valves prior to reactor restart An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to identify and correct a condition adverse to quality in a timely manner. Specifically, following the reactor scram on October 19, 2006, licensee personnel failed to properly evaluate discrepancies between the expected response of Feedwater Isolation Valves FWS-MOV7A and FWS-MOV7B, operator observation of valve indication, and indication of actual plant parameters affected by the valves, prior to restarting the reactor on October 22, 2006.
This violation was greater than minor because it was associated with the problem identification and resolution and the human performance attributes of the initiating events cornerstone and affected 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. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones,"
because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the violation only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program that ensured timely resolution of conditions adverse to quality.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of reactor operators to perform an adequate control board walkdown resulting in failure to identify that feedwater isolation valves were closing A self-revealing noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed conduct-of-operations procedures. Specifically, on October 19, 2006, two senior reactor operators (one on-coming and one off-going), conducting turnover activities, and the at-the-controls reactor operator failed to identify that the push buttons for Main Feedwater Isolation Valves 7A and 7B were out of alignment upon panel inspection during panel walk downs conducted in accordance with Entergy Operations Procedure EN-OP-115, "Conduct of Operations," Revision 2.
This violation was greater than minor because it was associated with the human performance attribute of the initiating events cornerstone and affected 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. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was initially determined to have very low safety significance because the violation only increased the initiating event likelihood by a very small amount and
 
the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to effectively use human error prevention techniques, such as self and peer checking.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to identify, place in the corrective action program, and correct deficiencies with Chart Recorder C33-R608 prior to restarting the reactor An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to identify and correct a condition adverse to quality in a timely manner. Specifically, on October 19, 2006, a licensed reactor operator noted a nonconforming condition with Strip Chart Recorder C33-R608 following the fall of the chart paper mechanism and discussed this with his supervision. However, this condition was not documented in the condition reporting process, the recorder was not properly inspected and repaired by qualified maintenance technicians prior to reactor restart, and at least one member of the on-site safety review committee may have been misinformed about the extent and composition of the evaluation and repair activities conducted on control room recorders prior to authorizing plant restart on October 22, 2006.
This finding was greater than minor because it was associated with the problem identification and resolution and the human performance attributes 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 because the chart recorder was left in a condition that had resulted in a reactor scram. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this finding was determined to be of very low safety significance because it only impacted the plant for a 2-day period. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program with a low threshold for identifying issues.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to provide complete corrective actions to address the probable cause of the October 19, 2006, scram, prior to restarting the reactor An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to correct a condition adverse to quality. Specifically, following the reactor scram on October 19, 2006, licensee personnel determined that the probable cause of the scram was a human performance error while handling the chart recorder. However, while significant corrective actions were taken, these actions did not completely address this probable cause prior to restarting the reactor on October 22, 2006, in that, expectations for working over control panels were not fully conveyed.
This violation was greater than minor because it was associated with the problem identification and resolution and the human performance attributes 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 because expectations and/or guidance were not provided to licensed operators on how to correct paper take up problems on strip chart recorders while minimizing the risk of dropping components on controls. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would
 
be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to be of very low safety significance because it only impacted the plant for a limited period of time. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program that ensured timely resolution of conditions adverse to quality.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Mitigating Systems Significance:      Feb 28, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to place the reactor mode switch in the SHUTDOWN position following a reactor scram as required by abnormal operating procedures A self-revealing noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures. Specifically, the at-the-controls operator failed to perform an immediate action required by Abnormal Operating Procedure AOP-0001, "Reactor Scram," Revision 22, which required him to place the mode switch in the SHUTDOWN position. The failure to reposition the mode switch resulted in an inadvertent main steam isolation, complicating the scram recovery.
This violation was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system. A Phase 2 estimation was required because this violation represented a loss of function of the steam side of the power conversion system as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones." Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the errors only impacted the plant for a short period of time and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to effectively use human error prevention techniques.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:      Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify that the reactor mode switch was in the SHUTDOWN position following a reactor scram as required by emergency operating procedures An NRC-identified noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures. Specifically, the control room supervisor failed to follow Emergency Operating Procedure EOP-0001, "Reactor Pressure Vessel Control,"
Revision 20, which required him to verify that the mode switch was in the SHUTDOWN position. The failure to reposition the mode switch resulted in an inadvertent main steam isolation, complicating the scram recovery.
This violation was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system. A Phase 2 estimation was required because this violation represented a loss of function of the steam side of the power conversion system, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones." Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the errors only impacted the plant for a short period of time and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to provide
 
adequate management oversight in this situation.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:      Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to permit the safety/relief valves to cycle in automatic and to manually operate the safety/relief valves without driving level outside the prescribed level band as required by AOPs An NRC-identified noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures. Specifically, licensed operators operated the safety/relief valves manually contrary to Abnormal Operating Procedure AOP-0001, OSP-0053, Attachment 1B, "Post Scram Pressure Control Strategies," Revision 5, requirements to operate them in automatic with the main steam isolation valves closed. Additionally, operators failed to manually operate the safety/relief valves, as required, to control pressure in the prescribed pressure band, without driving level outside the prescribed level band.
This violation was more than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system because manual actions affect licensed operator capability to perform simultaneous actions. Using the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," the finding was of very low safety significance because it did not represent a loss of safety function nor did it screen as potentially significant to external initiators. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the effectiveness of communicating expectations regarding procedural compliance.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:      Feb 28, 2007 Identified By: NRC Item Type: FIN Finding Senior reactor operator relieved the watch during a transient without waiting for the plant to be in a stable condition, resulting in an inadvertent main steam isolation The team identified a finding for the failure of licensed operators to accomplish activities affecting quality in accordance with the standards established in the conduct-of-operations procedures. Specifically, on October 19, 2006, the on-coming control room supervisor relieved the watch during the loss of feedwater transient, instead of waiting for the plant to be in a stable condition, a self-imposed standard documented in Entergy Operations Procedure EN-OP-115, "Conduct of Operations," Revision 2. Although licensee personnel stated that turnover activities were essentially complete at the time, changing the watch at this time caused the at-the-controls reactor operator and other control room personnel to misunderstand who was in charge of the event response and contributed to the at-the-controls operator not placing the mode switch in the SHUTDOWN position, as required by Procedure AOP-0001, "Reactor Scram," Revision 22. The failure to reposition the mode switch resulted in an inadvertent main steam isolation.
This finding was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system, namely the main feedwater system. A Phase 2 estimation was required because this finding resulted in a loss of function of the steam side of the power conversion system as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones." Using the appropriate plant-specific Phase 2 worksheets, this finding was determined to have very low safety significance because the finding only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This finding has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to implement the roles and responsibilities of the senior reactor operators in the main control room as designed.
The licensee entered this performance deficiency into their corrective action program for resolution.
 
Inspection Report# : 2006013 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to identify a degraded condition of steam leak detection system Transmitter E31-N084B A self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified involving the failure to identify a degraded condition affecting the steam leak detection and Division II isolation logic for residual heat removal/reactor core isolation cooling systems. The degraded condition resulted in a spurious isolation of the reactor core isolation cooling system during power operations on November 23, 2006. This issue was entered into the licensees corrective action program as CR-RBS-2006-04460.
The finding was more than minor because it is associated with the mitigating system cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Manual Chapter 0609, "Significance Determination Process," a Phase 2 analysis concluded that the finding was of very low safety significance.
The cause of the finding is related to the crosscutting aspect of problem identification and resolution in that the licensee failed to completely and accurately identify the condition that caused a previous isolation of the reactor core isolation cooling system on October 1, 2004. This failure resulted in the spurious reactor core isolation cooling system isolation on November 23, 2006.
Inspection Report# : 2006005 (pdf)
Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding Inadequate procedure for reassembly of the turbine bypass valve hydraulic system filter cartridge A self-revealing finding of very low safety significance was reviewed involving an inadequate procedure for conducting maintenance on the turbine bypass valve hydraulic system filter cartridge. This resulted in the improper reassembly of the filter. The resultant hydraulic oil leak caused the main turbine bypass valves to be inoperable, and a power reduction to less than 23.8 percent power was required by Technical Specifications. This issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2006-02632.
The performance deficiency associated with this finding was: (1) the failure to provide adequate instructions for reassembly of the turbine bypass valve hydraulic system filter cartridge to ensure that the cover gasket was properly installed, and (2) the failure to perform an adequate operational leak test of the system. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and it affected the associated cornerstone objective to ensure the availability and reliability of a system that responds to initiating events to prevent undesirable consequences. The inspectors performed a Phase 2 analysis using Manual Chapter 0609 and determined that the finding was of very low safety significance. The cause of the finding was related to the crosscutting element of human performance in that the licensee failed to provide complete, accurate, up-to-date instructions in the maintenance work package to change the hydraulic oil filter cartridge.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to identify Division III ESF bus supply breaker not racked in A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was reviewed involving the failure of the licensee to identify that the normal supply breaker to the Division III 4.16 kV engineered safety features bus was not properly racked in for a period of 24 days following maintenance. This issue was entered into the licensee's corrective action program as CR-RBS-2006-02402.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of configuration control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609,
 
"Significance Determination Process," a Phase 3 analysis concluded that the finding was of very low safety significance.
The cause of the finding was related to the crosscutting aspect of problem identification and resolution in that the licensee failed to properly evaluate available indications to identify that the breaker was not properly racked in.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to adequately manage an increase in plant risk An NRC identified noncited violation of 10 CFR 50.65 Maintenance Rule Section (a)(4) was identified for the failure of the licensee to provide prescribed compensatory measures for two Orange shutdown risk conditions during Refueling Outage 13. Specifically, the preoutage risk assessment recommended that two work orders be in place for maintenance electricians to provide power to one spent fuel pool cooling pump in the event of problems with the running pump during periods of electrical bus maintenance. The inspectors found that the work packages were not in place before entering shutdown risk condition Orange on April 26, 2006, during the Division II engineering safety features bus testing, and May 3, 2006, during the Division I engineered safety features bus outage. This issue was entered into the licensee's corrective action program as CR-RBS-2006-01937.
The finding was more than minor because the licensee failed to implement a prescribed compensatory measure during the highest risk condition of Refueling Outage 13. The specific compensatory measures were called for in the preoutage risk assessment and the shutdown operations protection plan. The finding affected the mitigating system cornerstone because of the increased risk of a sustained loss of spent fuel pool cooling during core offloading operations. The finding could not be evaluated using the significance determination process, therefore the finding was reviewed by regional management and determined to be of very low safety significance. Factors that were considered included: (1) electrical maintenance technicians had previously performed the task of providing alternate power to a spent fuel pool cooling pump, (2) the necessary equipment was staged as part of the abnormal operating procedure for loss of decay heat removal, and (3) the relatively long time to boil of the spent fuel storage pool at that time during the refueling outage. The cause of the finding was related to the crosscutting aspect of human performance because the licensees planned maintenance activities and the predetermined increase in outage risk was not effectively managed by prescribed compensatory measures.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure to verify required offsite power breaker alignment An NRC identified noncited violation of Technical Specification 5.4.1.a was identified for the failure of the licensee to provide an adequate surveillance test procedure to perform Technical Specification Surveillance Requirement 3.8.1.1.
Specifically, STP-000-0102, Power Distribution Alignment Check, Revision 4, did not verify the required offsite power circuit breaker alignment and indicated power availability for the Division III 4.16 kV engineered safety features bus as required in Modes 1, 2, and 3. This issue was entered into the licensee's corrective action program as CR-RBS-2006-02675 and -02402.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of configuration control and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609, "Significance Determination Process," a Phase 3 analysis concluded that the finding was of very low safety significance.
Inspection Report# : 2006003 (pdf)
Barrier Integrity Significance:      Dec 31, 2006 Identified By: Self-Revealing
 
Item Type: NCV NonCited Violation Inadequate work instructions result in isolation of annulus pressure control system and automatic start of the Division II standby gas treatment system A self-revealing, noncited violation of Technical Specification 5.4.1.a was identified involving the failure to provide adequate maintenance instructions for replacement of relays in the Division I standby gas treatment system initiation logic.
As a result, on November 21, 2006, during relay replacement, the annulus pressure control system tripped and the Division II standby gas treatment system automatically initiated. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04445.
This finding was more than minor because it is associated with the barrier integrity cornerstone attribute of human performance affecting the cornerstone objective to provide reasonable assurance that the secondary containment barrier protects the public from radionuclide releases caused by accidents and events. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because only the standby gas treatment system was affected. The cause of the finding is related to the crosscutting element of human performance in that the licensee failed to provide complete, accurate, and up-to-date instructions in the work package to replace the relays in the Division I standby gas treatment system initiation logic.
Inspection Report# : 2006005 (pdf)
Significance:      Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow procedure resulted in loss of power to safety-related instrumentation bus and isolation of reactor water cleanup system A self-revealing, noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow Procedure SOP-0048, 120 Vac System, Revision 303. Due to ineffective self- and peer-checking a procedure step was missed, resulting in inadvertent isolation of the reactor water cleanup and the suppression pool cooling and cleanup systems. This issue was entered into the licensee's corrective action program as CR-RBS-2006-03874.
The finding was more than minor because the loss of the reactor water cleanup system, providing reactor water chemistry control, affects the fuel barrier integrity cornerstone attribute of configuration control. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because only the fuel cladding barrier was affected. The cause of the finding is related to the crosscutting element of human performance in that operations personnel failed to make proper use of human performance techniques of self- and peer-checking.
Inspection Report# : 2006005 (pdf)
Significance:      Dec 31, 2006 Identified By: Self-Revealing Item Type: FIN Finding Newly installed reactor water cleanup pump coupling failed because it was beyond its expected service lifetime A self-revealing finding was identified involving the installation of a pump coupling that exceeded vendor shelf- and service-life recommendations on November 15, 2006. As a result, the reactor water cleanup Pump A coupling failed on November 28, 2006, requiring operators to remove from service the reactor water cleanup pump and a demineralizer affecting the primary means of reactor water chemistry control. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04488 and -04517.
The finding is greater than minor because it would become a more significant safety concern if left uncorrected, since failure of similar couplings affecting other plant components, such as the drywell floor and equipment drain pumps, would require a plant shutdown to make repairs. The finding affected the barrier integrity cornerstone. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because the finding only affected the fuel cladding barrier.
Inspection Report# : 2006005 (pdf)
 
Emergency Preparedness Significance:      May 10, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain a Standard Scheme of Emergency Classification and Action Levels in Use The NRC has considered the information developed during the inspection, the EOI position on the issue which was attached to the inspection report, the information you provided at the Regulatory Conference, and the information provided by your staff in a {{letter dated|date=July 28, 2006|text=July 28, 2006, letter}} following the conference. On the basis of this information, the NRC has concluded that a violation occurred. The violation involves a failure to meet 10 CFR 50.54(q), which requires that the licensee follow and maintain in effect emergency plans which meet the standards in 10 CFR 50.47(b). Specifically, during periods when certain seismic monitoring instrumentation was out of service, the licensee could not implement the emergency actions levels as described in the applicable Emergency Plan implementing procedure. However, there was other seismic instrumentation available during these periods that could be used to determine the ground force acceleration associated with a seismic event in the vicinity of the River Bend Station. This information could then be used by the Operations Shift Manager or Emergency Director to determine the correct classification for a seismic event; although, the classification could be delayed for as long as 4 hours. On the basis of this information, the NRC has concluded that the inspection finding did not represent a degradation of a risk significant planning standard function, as defined in Appendix B of NRCs Inspection Manual Chapter 0609, and therefore is of very low safety significance.
Also, this finding had crosscutting aspects in the area of problem identification and resolution because the River Bend Station staff did not identify the effect that inoperable seismic monitoring instrumentation had on the ability to implement the River Bend Station Emergency Plan and did not effectively utilize pertinent industry operating experience to prevent the performance deficiency.
Inspection Report# : 2006005 (pdf)
Inspection Report# : 2006011 (pdf)
Occupational Radiation Safety Significance:      Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Licensee failed to perform a radiological survey in off-gas sample room after radiological conditions had changed A self-revealing, noncited violation of 10 CFR 20.1501(a)(2) was identified involving the failure of radiation protection personnel to perform a survey in the off-gas sample room during main condenser leak testing. As a result, when a chemistry technician entered the room to obtain a grab sample, his electronic alarming dosimeter alarmed unexpectedly.
When another chemistry technician reached into the room to perform a survey of the test equipment, his dosimeter also alarmed. It was later determined that they were exposed to a dose rate of 440 and 521 millirem per hour, respectively. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04340.
The finding was more than minor because it was associated with the occupational radiation safety cornerstone attribute of programs and processes, such as the monitoring of radiological conditions, specifically the failure to perform a survey following changes in radiological conditions in the off-gas sample room, and affects the associated cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Utilizing Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for an overexposure, or (4) an impaired ability to assess dose. The cause of the finding was related to the crosscutting element of problem identification and resolution in that the licensee failed to communicate to affected personnel in a timely manner internal operating experience, specifically, while there was off-gas flow through the condenser leak test equipment, radiological conditions would increase.
Inspection Report# : 2006005 (pdf)
 
Significance:      Sep 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow radiation work permit requirements The inspector reviewed a self-revealing, noncited violation of Technical Specification 5.4.1 resulting from workers' failure to follow radiation work permit requirements. Two workers performing a scaffolding modification in the inclined fuel transfer system canal became externally and internally contaminated. As the workers were exiting the controlled access area, they alarmed the personnel contamination monitors. Based upon the whole-body count results, the licensee assigned a committed effective dose equivalent of 30 millirem to one worker and 70 millirem to the other worker. The licensee's investigation determined that the workers did not inform radiation protection personnel that they would be lowering the scaffolding 3 feet below surveyed areas and contamination control devices. Consequently, the workers were in radiological conditions not bounded by the radiation work permit and as low as is reasonably achievable planners did not have a chance to conduct a total effective dose equivalent as low as is reasonably achievable review to determine if respiratory protection was necessary. As a corrective action, the licensee is incorporating a lessons learned item associated with this event into radiation worker training.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure/contamination control) and affects the Occupational Radiation Safety cornerstone objective in that the failure to follow radiation work permit instructions resulted in additional personnel exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because the workers failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2006004 (pdf)
Significance:      Sep 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to evaluate radiological conditions The inspector reviewed a self-revealing, noncited violation of 10 CFR 20.1501(a) resulting from the licensee's failure to correctly measure the airborne radioactivity where personnel worked. The licensee's review of the January 26, 2006, contamination event identified that the air sample taken to support the work activity was positioned above the high-efficiency particulate air hose suction in an air flow area above the actual work area. This meant that the air sample was not representative of the workers' actual work area. In addition, the radiation protection technician providing continuous job coverage failed to identify the deficiency and adjust the position of the air sampler. As a corrective action, the licensee is incorporating a lessons learned item associated with this event into the radiation protection technician training.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affects the Occupational Radiation Safety cornerstone objective in that an inadequate evaluation of the hazards could lead to inadequate radiation protection and dose saving measures. This finding could also be reasonably viewed as a precursor to a significant event, such as a personnel overexposure, had contamination levels been higher. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because radiation protection personnel failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2006004 (pdf)
Significance:      Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to evaluate radiological conditions The inspector identified a noncited violation of 10 CFR 20.1501(a), resulting from the licensees use of an inadequate alpha contamination survey technique. The inspector determined that the licensee's procedure for the use of the Eberline SAC-4 alpha scintillation counter established a screening limit that did not allow sufficient sample activity for the discovery of
 
alpha emitting radionuclides. Therefore, the inspector concluded that surveys using this technique could not identify alpha contamination and were inadequate. As a corrective action, the licensee adapted the corporate procedural guidance, which raised the maximum sample activity.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affects the Occupational Radiation Safety cornerstone objective in that an inadequate evaluation of the hazards could lead to inadequate radiation protection and dose saving measures. This finding could also be reasonably viewed as a precursor to a significant event, such as a personnel overexposure, had contamination levels been higher. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because the licensee used procedures that were inadequate to ensure that alpha contamination was identified.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to control access to a high radiation area The inspector reviewed a self-revealing noncited violation of Technical Specification 5.7.1, resulting from the licensees failure to control access to a high radiation area. While transferring reverse osmosis system filters in the radwaste building, the licensee allowed two workers to inadvertently enter a high radiation area. This occurred after a guard prematurely left his post in front of the 123 foot elevation elevator door. The highest dose rate recorded by an electronic alarming dosimeter was 164 millirem per hour. The guard returned and evacuated the workers before they accrued additional radiation dose.
Planned corrective action was still being evaluated by the licensee at the conclusion of the inspection.
The finding was more than minor because it was associated with the occupational radiation safety attribute of exposure control and affected the cornerstone objective in that not controlling a high radiation area could increase personal exposure.
Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance in that the failure of the individual to guard the elevator door directly contributed to the violation.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform airborne radiation survey The inspector identified a noncited violation of 10 CFR 20.1501(a) because the licensee failed to survey airborne radioactivity. During the removal of local power range monitors, the licensee started collecting an air sample of the work area, but discarded the sample before analyzing it. Successful passage through the portal monitors at the exit of the controlled access area confirmed that no worker experienced an uptake of radioactive material. Planned corrective action is still being evaluated.
The finding was more than minor because it was associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance in that the failure to maintain the sample for analysis directly contributed to the violation.
Inspection Report# : 2006003 (pdf)
 
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : June 01, 2007
 
River Bend 1 2Q/2007 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Maintenance Instructions A self-revealing finding was identified involving the failure of maintenance personnel to follow maintenance instructions resulting in the failure to properly seal the desiccant retention strainer of an instrument air dryer. As a result, desiccant was released from the dryer tower and became lodged in an outlet shuttle valve causing it to stick open that resulted in lowering the instrument air header pressure. This condition caused operators to enter the abnormal operating procedure for loss of instrument air, an automatic start of standby air compressors, and the automatic cross-connect of service air to the instrument air header. These actions restored instrument air pressure preventing a significant plant transient. This issue was entered into the licensee's corrective action program as CR-RBS-2007-00438.
The finding was more than minor because it would become a more significant safety concern if left uncorrected in that an air dryer failure could result in a complete loss of instrument air. The finding affected the initiating event cornerstone. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because there was no actual loss of instrument air. The cause of the finding was related to the crosscutting element of human performance in that the maintenance technicians failed to properly self and peer check the adequacy of the retention strainer seal during maintenance of instrument air Dryer 2 on January 12, 2007. As a result, desiccant was released causing an outlet shuttle valve to stick open.
Inspection Report# : 2007002 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct discrepancies between the design function and observed response of the feedwater isolation valves prior to reactor restart An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to identify and correct a condition adverse to quality in a timely manner.
Specifically, following the reactor scram on October 19, 2006, licensee personnel failed to properly evaluate discrepancies between the expected response of Feedwater Isolation Valves FWS-MOV7A and FWS-MOV7B, operator observation of valve indication, and indication of actual plant parameters affected by the valves, prior to restarting the reactor on October 22, 2006.
This violation was greater than minor because it was associated with the problem identification and resolution and the human performance attributes of the initiating events cornerstone and affected 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. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the violation only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program that ensured timely resolution of conditions adverse to quality.
 
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of reactor operators to perform an adequate control board walkdown resulting in failure to identify that feedwater isolation valves were closing A self-revealing noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed conduct-of-operations procedures. Specifically, on October 19, 2006, two senior reactor operators (one on-coming and one off-going),
conducting turnover activities, and the at-the-controls reactor operator failed to identify that the push buttons for Main Feedwater Isolation Valves 7A and 7B were out of alignment upon panel inspection during panel walk downs conducted in accordance with Entergy Operations Procedure EN-OP-115, "Conduct of Operations," Revision 2.
This violation was greater than minor because it was associated with the human performance attribute of the initiating events cornerstone and affected 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. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was initially determined to have very low safety significance because the violation only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to effectively use human error prevention techniques, such as self and peer checking.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to identify, place in the corrective action program, and correct deficiencies with Chart Recorder C33-R608 prior to restarting the reactor An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to identify and correct a condition adverse to quality in a timely manner.
Specifically, on October 19, 2006, a licensed reactor operator noted a nonconforming condition with Strip Chart Recorder C33-R608 following the fall of the chart paper mechanism and discussed this with his supervision.
However, this condition was not documented in the condition reporting process, the recorder was not properly inspected and repaired by qualified maintenance technicians prior to reactor restart, and at least one member of the on-site safety review committee may have been misinformed about the extent and composition of the evaluation and repair activities conducted on control room recorders prior to authorizing plant restart on October 22, 2006.
This finding was greater than minor because it was associated with the problem identification and resolution and the human performance attributes 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 because the chart recorder was left in a condition that had resulted in a reactor scram. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this finding was determined to be of very low safety significance because it only impacted the plant for a 2-day period. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program with a low threshold for identifying issues.
 
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to provide complete corrective actions to address the probable cause of the October 19, 2006, scram, prior to restarting the reactor An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to correct a condition adverse to quality. Specifically, following the reactor scram on October 19, 2006, licensee personnel determined that the probable cause of the scram was a human performance error while handling the chart recorder. However, while significant corrective actions were taken, these actions did not completely address this probable cause prior to restarting the reactor on October 22, 2006, in that, expectations for working over control panels were not fully conveyed.
This violation was greater than minor because it was associated with the problem identification and resolution and the human performance attributes 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 because expectations and/or guidance were not provided to licensed operators on how to correct paper take up problems on strip chart recorders while minimizing the risk of dropping components on controls. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to be of very low safety significance because it only impacted the plant for a limited period of time. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program that ensured timely resolution of conditions adverse to quality.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Mitigating Systems Significance:        Apr 27, 2007 Identified By: NRC Item Type: FIN Finding Foreign Material Found in Residual Heat Removal Room Sump Pump Discharge Check Valve The team identified a finding because the licensee failed to address control of foreign material in the Train B residual heat removal room in June 2003. Consequently, on March 5, 2007, maintenance technicians found foreign material in one of the sump pump discharge check valves. This failure to control foreign material resulted in sump high level alarms, which had caused the operators to enter the emergency operating procedure for auxiliary building room flooding on three different occasions. The licensee documented this deficiency in Condition Report 2007-00859.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability of the residual heat removal system. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because there was no actual loss of the residual heat removal system function and it did not screen as potentially risk significant for an internal flooding event. The cause of the finding was related to the crosscutting element of human performance work practices in that licensee management failed to communicate and enforce compliance with the site foreign material control program.
Inspection Report# : 2007009 (pdf)
 
Significance:      Mar 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Implement Freeze Protection Compensatory Measures The inspectors identified a finding involving the failure of operators to implement compensatory measures for cold weather conditions when a ventilation heater for a safety related standby cooling tower pipe chase was out of service during the winters from 2003 through 2006. This issue was entered into the licensees corrective action program as CR-RBS-2007-00399.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of equipment performance and affected the associated cornerstone objective. The finding was determined to have very low safety significance because it did not result in a actual loss of the standby service water system and it was determined by a Phase 3 analysis not to be risk significant due to external events. The cause of the finding was related to the crosscutting aspect of problem identification and resolution in that the licensee failed to identify that freeze protection equipment in the area was out of service each winter from 2003 through 2006 requiring compensatory measures.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct a Degraded Residual Heat Removal System Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified involving the failure to promptly identify and correct a condition adverse to quality. Specifically, on August 3, 2005, residual heat removal Train A fuel pool cooling assist Valve E12-MOVF037A failed to fully close during actuation. The failure to correct the problem resulted in recurrence of the valve failing to fully close on April 11, 2006, and January 7, 2007. This issue was entered into the licensee's corrective action program as CR-RBS-2006-01326.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and affected the associated cornerstone objective. The finding had very low safety significance because it did not represent a loss of the residual heat removal system safety function. The cause of the finding was related to the crosscutting element of problem identification and resolution in that the licensee did not thoroughly evaluate the problem such that the resolution would address the cause of the failure of Valve E12-MOVF037A to fully close on August 3, 2005.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Instructions for Installation of a Compression Fitting The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings, for the licensees failure to provide adequate work instructions for repairing a failed tubing compression fitting on the Division I emergency diesel generator jacket cooling water system. Specifically, the repair inappropriately had tubing entering a compression fitting at an angle that could result in failure as had previously been encountered on the same fitting. This issue was entered into the licensee's corrective action program as CR-RBS-2007-01496.
The finding was more than minor because it would become a more significant event if left uncorrected in that failure to install and repair tubing fittings correctly can lead to subsequent failure. The finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because the repair did not result in an actual loss of function of the Division I emergency diesel generator. The cause of the finding was related to the crosscutting element of human performance in that the licensee did not effectively communicate expectations for proper assembly of tubing fittings on safety related equipment.
 
Inspection Report# : 2007002 (pdf)
Significance:      Feb 28, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to place the reactor mode switch in the SHUTDOWN position following a reactor scram as required by abnormal operating procedures A self-revealing noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures. Specifically, the at-the-controls operator failed to perform an immediate action required by Abnormal Operating Procedure AOP-0001, "Reactor Scram," Revision 22, which required him to place the mode switch in the SHUTDOWN position. The failure to reposition the mode switch resulted in an inadvertent main steam isolation, complicating the scram recovery.
This violation was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system. A Phase 2 estimation was required because this violation represented a loss of function of the steam side of the power conversion system as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones."
Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the errors only impacted the plant for a short period of time and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to effectively use human error prevention techniques.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:      Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify that the reactor mode switch was in the SHUTDOWN position following a reactor scram as required by emergency operating procedures An NRC-identified noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures.
Specifically, the control room supervisor failed to follow Emergency Operating Procedure EOP-0001, "Reactor Pressure Vessel Control," Revision 20, which required him to verify that the mode switch was in the SHUTDOWN position. The failure to reposition the mode switch resulted in an inadvertent main steam isolation, complicating the scram recovery.
This violation was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system. A Phase 2 estimation was required because this violation represented a loss of function of the steam side of the power conversion system, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones."
Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the errors only impacted the plant for a short period of time and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to provide adequate management oversight in this situation.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:      Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation
 
Operators failed to permit the safety/relief valves to cycle in automatic and to manually operate the safety/relief valves without driving level outside the prescribed level band as required by AOPs An NRC-identified noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures.
Specifically, licensed operators operated the safety/relief valves manually contrary to Abnormal Operating Procedure AOP-0001, OSP-0053, Attachment 1B, "Post Scram Pressure Control Strategies," Revision 5, requirements to operate them in automatic with the main steam isolation valves closed. Additionally, operators failed to manually operate the safety/relief valves, as required, to control pressure in the prescribed pressure band, without driving level outside the prescribed level band.
This violation was more than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system because manual actions affect licensed operator capability to perform simultaneous actions. Using the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," the finding was of very low safety significance because it did not represent a loss of safety function nor did it screen as potentially significant to external initiators. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the effectiveness of communicating expectations regarding procedural compliance.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: FIN Finding Senior reactor operator relieved the watch during a transient without waiting for the plant to be in a stable condition, resulting in an inadvertent main steam isolation The team identified a finding for the failure of licensed operators to accomplish activities affecting quality in accordance with the standards established in the conduct-of-operations procedures. Specifically, on October 19, 2006, the on-coming control room supervisor relieved the watch during the loss of feedwater transient, instead of waiting for the plant to be in a stable condition, a self-imposed standard documented in Entergy Operations Procedure EN-OP-115, "Conduct of Operations," Revision 2. Although licensee personnel stated that turnover activities were essentially complete at the time, changing the watch at this time caused the at-the-controls reactor operator and other control room personnel to misunderstand who was in charge of the event response and contributed to the at-the-controls operator not placing the mode switch in the SHUTDOWN position, as required by Procedure AOP-0001, "Reactor Scram," Revision 22. The failure to reposition the mode switch resulted in an inadvertent main steam isolation.
This finding was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system, namely the main feedwater system. A Phase 2 estimation was required because this finding resulted in a loss of function of the steam side of the power conversion system as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones." Using the appropriate plant-specific Phase 2 worksheets, this finding was determined to have very low safety significance because the finding only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This finding has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to implement the roles and responsibilities of the senior reactor operators in the main control room as designed.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to identify a degraded condition of steam leak detection system Transmitter E31-N084B A self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was
 
identified involving the failure to identify a degraded condition affecting the steam leak detection and Division II isolation logic for residual heat removal/reactor core isolation cooling systems. The degraded condition resulted in a spurious isolation of the reactor core isolation cooling system during power operations on November 23, 2006. This issue was entered into the licensees corrective action program as CR-RBS-2006-04460.
The finding was more than minor because it is associated with the mitigating system cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Manual Chapter 0609, "Significance Determination Process," a Phase 2 analysis concluded that the finding was of very low safety significance. The cause of the finding is related to the crosscutting aspect of problem identification and resolution in that the licensee failed to completely and accurately identify the condition that caused a previous isolation of the reactor core isolation cooling system on October 1, 2004. This failure resulted in the spurious reactor core isolation cooling system isolation on November 23, 2006.
Inspection Report# : 2006005 (pdf)
Significance:        Sep 30, 2006 Identified By: Self-Revealing Item Type: FIN Finding Inadequate procedure for reassembly of the turbine bypass valve hydraulic system filter cartridge A self-revealing finding of very low safety significance was reviewed involving an inadequate procedure for conducting maintenance on the turbine bypass valve hydraulic system filter cartridge. This resulted in the improper reassembly of the filter. The resultant hydraulic oil leak caused the main turbine bypass valves to be inoperable, and a power reduction to less than 23.8 percent power was required by Technical Specifications. This issue was entered into the licensee's corrective action program as Condition Report CR-RBS-2006-02632.
The performance deficiency associated with this finding was: (1) the failure to provide adequate instructions for reassembly of the turbine bypass valve hydraulic system filter cartridge to ensure that the cover gasket was properly installed, and (2) the failure to perform an adequate operational leak test of the system. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and it affected the associated cornerstone objective to ensure the availability and reliability of a system that responds to initiating events to prevent undesirable consequences. The inspectors performed a Phase 2 analysis using Manual Chapter 0609 and determined that the finding was of very low safety significance. The cause of the finding was related to the crosscutting element of human performance in that the licensee failed to provide complete, accurate, up-to-date instructions in the maintenance work package to change the hydraulic oil filter cartridge.
Inspection Report# : 2006004 (pdf)
Barrier Integrity Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate work instructions result in isolation of annulus pressure control system and automatic start of the Division II standby gas treatment system A self-revealing, noncited violation of Technical Specification 5.4.1.a was identified involving the failure to provide adequate maintenance instructions for replacement of relays in the Division I standby gas treatment system initiation logic. As a result, on November 21, 2006, during relay replacement, the annulus pressure control system tripped and the Division II standby gas treatment system automatically initiated. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04445.
This finding was more than minor because it is associated with the barrier integrity cornerstone attribute of human performance affecting the cornerstone objective to provide reasonable assurance that the secondary containment barrier protects the public from radionuclide releases caused by accidents and events. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety
 
significance because only the standby gas treatment system was affected. The cause of the finding is related to the crosscutting element of human performance in that the licensee failed to provide complete, accurate, and up-to-date instructions in the work package to replace the relays in the Division I standby gas treatment system initiation logic.
Inspection Report# : 2006005 (pdf)
Significance:      Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow procedure resulted in loss of power to safety-related instrumentation bus and isolation of reactor water cleanup system A self-revealing, noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow Procedure SOP-0048, 120 Vac System, Revision 303. Due to ineffective self- and peer-checking a procedure step was missed, resulting in inadvertent isolation of the reactor water cleanup and the suppression pool cooling and cleanup systems. This issue was entered into the licensee's corrective action program as CR-RBS-2006-03874.
The finding was more than minor because the loss of the reactor water cleanup system, providing reactor water chemistry control, affects the fuel barrier integrity cornerstone attribute of configuration control. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because only the fuel cladding barrier was affected. The cause of the finding is related to the crosscutting element of human performance in that operations personnel failed to make proper use of human performance techniques of self- and peer-checking.
Inspection Report# : 2006005 (pdf)
Significance:      Dec 31, 2006 Identified By: Self-Revealing Item Type: FIN Finding Newly installed reactor water cleanup pump coupling failed because it was beyond its expected service lifetime A self-revealing finding was identified involving the installation of a pump coupling that exceeded vendor shelf- and service-life recommendations on November 15, 2006. As a result, the reactor water cleanup Pump A coupling failed on November 28, 2006, requiring operators to remove from service the reactor water cleanup pump and a demineralizer affecting the primary means of reactor water chemistry control. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04488 and -04517.
The finding is greater than minor because it would become a more significant safety concern if left uncorrected, since failure of similar couplings affecting other plant components, such as the drywell floor and equipment drain pumps, would require a plant shutdown to make repairs. The finding affected the barrier integrity cornerstone. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because the finding only affected the fuel cladding barrier.
Inspection Report# : 2006005 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Licensee failed to perform a radiological survey in off-gas sample room after radiological conditions had changed A self-revealing, noncited violation of 10 CFR 20.1501(a)(2) was identified involving the failure of radiation
 
protection personnel to perform a survey in the off-gas sample room during main condenser leak testing. As a result, when a chemistry technician entered the room to obtain a grab sample, his electronic alarming dosimeter alarmed unexpectedly. When another chemistry technician reached into the room to perform a survey of the test equipment, his dosimeter also alarmed. It was later determined that they were exposed to a dose rate of 440 and 521 millirem per hour, respectively. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04340.
The finding was more than minor because it was associated with the occupational radiation safety cornerstone attribute of programs and processes, such as the monitoring of radiological conditions, specifically the failure to perform a survey following changes in radiological conditions in the off-gas sample room, and affects the associated cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Utilizing Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for an overexposure, or (4) an impaired ability to assess dose.
The cause of the finding was related to the crosscutting element of problem identification and resolution in that the licensee failed to communicate to affected personnel in a timely manner internal operating experience, specifically, while there was off-gas flow through the condenser leak test equipment, radiological conditions would increase.
Inspection Report# : 2006005 (pdf)
Significance:      Sep 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow radiation work permit requirements The inspector reviewed a self-revealing, noncited violation of Technical Specification 5.4.1 resulting from workers' failure to follow radiation work permit requirements. Two workers performing a scaffolding modification in the inclined fuel transfer system canal became externally and internally contaminated. As the workers were exiting the controlled access area, they alarmed the personnel contamination monitors. Based upon the whole-body count results, the licensee assigned a committed effective dose equivalent of 30 millirem to one worker and 70 millirem to the other worker. The licensee's investigation determined that the workers did not inform radiation protection personnel that they would be lowering the scaffolding 3 feet below surveyed areas and contamination control devices. Consequently, the workers were in radiological conditions not bounded by the radiation work permit and as low as is reasonably achievable planners did not have a chance to conduct a total effective dose equivalent as low as is reasonably achievable review to determine if respiratory protection was necessary. As a corrective action, the licensee is incorporating a lessons learned item associated with this event into radiation worker training.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure/contamination control) and affects the Occupational Radiation Safety cornerstone objective in that the failure to follow radiation work permit instructions resulted in additional personnel exposure. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because the workers failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2006004 (pdf)
Significance:      Sep 30, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to evaluate radiological conditions The inspector reviewed a self-revealing, noncited violation of 10 CFR 20.1501(a) resulting from the licensee's failure to correctly measure the airborne radioactivity where personnel worked. The licensee's review of the January 26, 2006, contamination event identified that the air sample taken to support the work activity was positioned above the high-efficiency particulate air hose suction in an air flow area above the actual work area. This meant that the air sample was not representative of the workers' actual work area. In addition, the radiation protection technician providing continuous job coverage failed to identify the deficiency and adjust the position of the air sampler. As a corrective action, the licensee is incorporating a lessons learned item associated with this event into the radiation
 
protection technician training.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affects the Occupational Radiation Safety cornerstone objective in that an inadequate evaluation of the hazards could lead to inadequate radiation protection and dose saving measures. This finding could also be reasonably viewed as a precursor to a significant event, such as a personnel overexposure, had contamination levels been higher. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because radiation protection personnel failed to use error prevention tools such as self- and peer-checking.
Inspection Report# : 2006004 (pdf)
Significance:      Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to evaluate radiological conditions The inspector identified a noncited violation of 10 CFR 20.1501(a), resulting from the licensees use of an inadequate alpha contamination survey technique. The inspector determined that the licensee's procedure for the use of the Eberline SAC-4 alpha scintillation counter established a screening limit that did not allow sufficient sample activity for the discovery of alpha emitting radionuclides. Therefore, the inspector concluded that surveys using this technique could not identify alpha contamination and were inadequate. As a corrective action, the licensee adapted the corporate procedural guidance, which raised the maximum sample activity.
This finding is greater than minor because it is associated with one of the cornerstone attributes (exposure control) and affects the Occupational Radiation Safety cornerstone objective in that an inadequate evaluation of the hazards could lead to inadequate radiation protection and dose saving measures. This finding could also be reasonably viewed as a precursor to a significant event, such as a personnel overexposure, had contamination levels been higher. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess doses. Additionally, this finding has a crosscutting aspect in the area of human performance because the licensee used procedures that were inadequate to ensure that alpha contamination was identified.
Inspection Report# : 2006004 (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 24, 2007
 
River Bend 1 3Q/2007 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Vendor Recommendations A self-revealing finding was identified involving the failure to implement 1998 vendor recommendations associated with the potential for vibration induced degradation of recirculation loop gate valves. This resulted in the failure to identify and implement timely corrective actions prior to disk to stem separation of recirculation Pump A discharge gate valve that occurred on May 21, 2007. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-02113.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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 the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have a very low safety significance because the finding did not contribute to the likelihood that mitigation equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007003 (pdf)
Significance:      Jun 30, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Work Instructions The inspectors identified a finding involving inadequate maintenance instructions for opening a stuck closed feedwater regulating Valve A isolation valve. Specifically, the instructions failed to account for the system being pressurized resulting in unexpected valve stem movement while technicians were removing the manual operator from the valve on June 10, 2007. This deficiency could have resulted in personnel harm or an unexpected and uncontrolled plant transient. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-02576.
The finding was more than minor because it could become a more significant safety concern if left uncorrected. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have very low safety significance because the deficiency did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. No violation of NRC requirements occurred. The cause of this finding was related to the human performance crosscutting component of resources because the licensee did not ensure a complete and accurate work package was available prior to the start of the job (H.2(c)).
Inspection Report# : 2007003 (pdf)
Significance:      Jun 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Install Scram Discharge Instrument Volume Vent Plug A self-revealing noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow a surveillance procedure for scram discharge instrument volume water level channel calibration. Specifically, on February 9, 2007, an instrument line plug was not replaced following surveillance testing. As a result, on May 5, 2007, following a reactor scram, reactor water sprayed out of the scram discharge instrument volume and
 
contaminated some accessible portions of the containment building causing three inadvertent personnel contamination events. This issue was entered into the licensees corrective action program as condition Report CR-RBS-2007-01809.
The finding was more than minor because it was associated with the initiating event cornerstone attribute of equipment performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency resulted in a reactor coolant leak greater than the Technical Specification limit for identified reactor coolant system leakage. Using the plant-specific Phase 2 risk-informed notebook, this violation was determined to have very low safety significance because the violation only increased the likelihood of a small-break loss of coolant accident by a very small amount and mitigation capability was unaffected. The cause of the finding was related to the human performance crosscutting component of work practices because neither self nor peer checking actions identified the failure to replace the vent plug (H.4(a)).
Inspection Report# : 2007003 (pdf)
Significance:        Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Maintenance Instructions A self-revealing finding was identified involving the failure of maintenance personnel to follow maintenance instructions resulting in the failure to properly seal the desiccant retention strainer of an instrument air dryer. As a result, desiccant was released from the dryer tower and became lodged in an outlet shuttle valve causing it to stick open that resulted in lowering the instrument air header pressure. This condition caused operators to enter the abnormal operating procedure for loss of instrument air, an automatic start of standby air compressors, and the automatic cross-connect of service air to the instrument air header. These actions restored instrument air pressure preventing a significant plant transient. This issue was entered into the licensee's corrective action program as CR-RBS-2007-00438.
The finding was more than minor because it would become a more significant safety concern if left uncorrected in that an air dryer failure could result in a complete loss of instrument air. The finding affected the initiating event cornerstone. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because there was no actual loss of instrument air. The cause of the finding was related to the crosscutting element of human performance in that the maintenance technicians failed to properly self and peer check the adequacy of the retention strainer seal during maintenance of instrument air Dryer 2 on January 12, 2007. As a result, desiccant was released causing an outlet shuttle valve to stick open.
Inspection Report# : 2007002 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct discrepancies between the design function and observed response of the feedwater isolation valves prior to reactor restart An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to identify and correct a condition adverse to quality in a timely manner.
Specifically, following the reactor scram on October 19, 2006, licensee personnel failed to properly evaluate discrepancies between the expected response of Feedwater Isolation Valves FWS-MOV7A and FWS-MOV7B, operator observation of valve indication, and indication of actual plant parameters affected by the valves, prior to restarting the reactor on October 22, 2006.
This violation was greater than minor because it was associated with the problem identification and resolution and the human performance attributes of the initiating events cornerstone and affected 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. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers
 
Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the violation only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program that ensured timely resolution of conditions adverse to quality.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of reactor operators to perform an adequate control board walkdown resulting in failure to identify that feedwater isolation valves were closing A self-revealing noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed conduct-of-operations procedures. Specifically, on October 19, 2006, two senior reactor operators (one on-coming and one off-going),
conducting turnover activities, and the at-the-controls reactor operator failed to identify that the push buttons for Main Feedwater Isolation Valves 7A and 7B were out of alignment upon panel inspection during panel walk downs conducted in accordance with Entergy Operations Procedure EN-OP-115, "Conduct of Operations," Revision 2.
This violation was greater than minor because it was associated with the human performance attribute of the initiating events cornerstone and affected 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. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was initially determined to have very low safety significance because the violation only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to effectively use human error prevention techniques, such as self and peer checking.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to identify, place in the corrective action program, and correct deficiencies with Chart Recorder C33-R608 prior to restarting the reactor An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to identify and correct a condition adverse to quality in a timely manner.
Specifically, on October 19, 2006, a licensed reactor operator noted a nonconforming condition with Strip Chart Recorder C33-R608 following the fall of the chart paper mechanism and discussed this with his supervision.
However, this condition was not documented in the condition reporting process, the recorder was not properly inspected and repaired by qualified maintenance technicians prior to reactor restart, and at least one member of the on-site safety review committee may have been misinformed about the extent and composition of the evaluation and repair activities conducted on control room recorders prior to authorizing plant restart on October 22, 2006.
This finding was greater than minor because it was associated with the problem identification and resolution and the human performance attributes 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 because the chart recorder was left in a condition that had resulted in a reactor scram. A Phase 2 estimation
 
was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this finding was determined to be of very low safety significance because it only impacted the plant for a 2-day period. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program with a low threshold for identifying issues.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to provide complete corrective actions to address the probable cause of the October 19, 2006, scram, prior to restarting the reactor An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to correct a condition adverse to quality. Specifically, following the reactor scram on October 19, 2006, licensee personnel determined that the probable cause of the scram was a human performance error while handling the chart recorder. However, while significant corrective actions were taken, these actions did not completely address this probable cause prior to restarting the reactor on October 22, 2006, in that, expectations for working over control panels were not fully conveyed.
This violation was greater than minor because it was associated with the problem identification and resolution and the human performance attributes 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 because expectations and/or guidance were not provided to licensed operators on how to correct paper take up problems on strip chart recorders while minimizing the risk of dropping components on controls. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to be of very low safety significance because it only impacted the plant for a limited period of time. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program that ensured timely resolution of conditions adverse to quality.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Mitigating Systems Significance:        Apr 27, 2007 Identified By: NRC Item Type: FIN Finding Foreign Material Found in Residual Heat Removal Room Sump Pump Discharge Check Valve The team identified a finding because the licensee failed to address control of foreign material in the Train B residual heat removal room in June 2003. Consequently, on March 5, 2007, maintenance technicians found foreign material in one of the sump pump discharge check valves. This failure to control foreign material resulted in sump high level alarms, which had caused the operators to enter the emergency operating procedure for auxiliary building room flooding on three different occasions. The licensee documented this deficiency in Condition Report 2007-00859.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of
 
equipment performance and affected the associated cornerstone objective to ensure the availability of the residual heat removal system. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because there was no actual loss of the residual heat removal system function and it did not screen as potentially risk significant for an internal flooding event. The cause of the finding was related to the crosscutting element of human performance work practices in that licensee management failed to communicate and enforce compliance with the site foreign material control program.
Inspection Report# : 2007009 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Implement Freeze Protection Compensatory Measures The inspectors identified a finding involving the failure of operators to implement compensatory measures for cold weather conditions when a ventilation heater for a safety related standby cooling tower pipe chase was out of service during the winters from 2003 through 2006. This issue was entered into the licensees corrective action program as CR-RBS-2007-00399.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of equipment performance and affected the associated cornerstone objective. The finding was determined to have very low safety significance because it did not result in a actual loss of the standby service water system and it was determined by a Phase 3 analysis not to be risk significant due to external events. The cause of the finding was related to the crosscutting aspect of problem identification and resolution in that the licensee failed to identify that freeze protection equipment in the area was out of service each winter from 2003 through 2006 requiring compensatory measures.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct a Degraded Residual Heat Removal System Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified involving the failure to promptly identify and correct a condition adverse to quality. Specifically, on August 3, 2005, residual heat removal Train A fuel pool cooling assist Valve E12-MOVF037A failed to fully close during actuation. The failure to correct the problem resulted in recurrence of the valve failing to fully close on April 11, 2006, and January 7, 2007. This issue was entered into the licensee's corrective action program as CR-RBS-2006-01326.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and affected the associated cornerstone objective. The finding had very low safety significance because it did not represent a loss of the residual heat removal system safety function. The cause of the finding was related to the crosscutting element of problem identification and resolution in that the licensee did not thoroughly evaluate the problem such that the resolution would address the cause of the failure of Valve E12-MOVF037A to fully close on August 3, 2005.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Instructions for Installation of a Compression Fitting The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings, for the licensees failure to provide adequate work instructions for repairing a failed tubing compression fitting on the Division I emergency diesel generator jacket cooling water system. Specifically, the repair inappropriately had tubing entering a compression fitting at an angle that could result in failure as had previously been encountered on the same fitting. This issue was entered into the licensee's corrective action program as CR-RBS-2007-01496.
 
The finding was more than minor because it would become a more significant event if left uncorrected in that failure to install and repair tubing fittings correctly can lead to subsequent failure. The finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because the repair did not result in an actual loss of function of the Division I emergency diesel generator. The cause of the finding was related to the crosscutting element of human performance in that the licensee did not effectively communicate expectations for proper assembly of tubing fittings on safety related equipment.
Inspection Report# : 2007002 (pdf)
Significance:      Feb 28, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to place the reactor mode switch in the SHUTDOWN position following a reactor scram as required by abnormal operating procedures A self-revealing noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures. Specifically, the at-the-controls operator failed to perform an immediate action required by Abnormal Operating Procedure AOP-0001, "Reactor Scram," Revision 22, which required him to place the mode switch in the SHUTDOWN position. The failure to reposition the mode switch resulted in an inadvertent main steam isolation, complicating the scram recovery.
This violation was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system. A Phase 2 estimation was required because this violation represented a loss of function of the steam side of the power conversion system as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones."
Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the errors only impacted the plant for a short period of time and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to effectively use human error prevention techniques.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:      Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify that the reactor mode switch was in the SHUTDOWN position following a reactor scram as required by emergency operating procedures An NRC-identified noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures.
Specifically, the control room supervisor failed to follow Emergency Operating Procedure EOP-0001, "Reactor Pressure Vessel Control," Revision 20, which required him to verify that the mode switch was in the SHUTDOWN position. The failure to reposition the mode switch resulted in an inadvertent main steam isolation, complicating the scram recovery.
This violation was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system. A Phase 2 estimation was required because this violation represented a loss of function of the steam side of the power conversion system, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones."
Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the errors only impacted the plant for a short period of time and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to provide adequate management oversight in this situation.
 
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to permit the safety/relief valves to cycle in automatic and to manually operate the safety/relief valves without driving level outside the prescribed level band as required by AOPs An NRC-identified noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures.
Specifically, licensed operators operated the safety/relief valves manually contrary to Abnormal Operating Procedure AOP-0001, OSP-0053, Attachment 1B, "Post Scram Pressure Control Strategies," Revision 5, requirements to operate them in automatic with the main steam isolation valves closed. Additionally, operators failed to manually operate the safety/relief valves, as required, to control pressure in the prescribed pressure band, without driving level outside the prescribed level band.
This violation was more than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system because manual actions affect licensed operator capability to perform simultaneous actions. Using the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," the finding was of very low safety significance because it did not represent a loss of safety function nor did it screen as potentially significant to external initiators. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the effectiveness of communicating expectations regarding procedural compliance.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: FIN Finding Senior reactor operator relieved the watch during a transient without waiting for the plant to be in a stable condition, resulting in an inadvertent main steam isolation The team identified a finding for the failure of licensed operators to accomplish activities affecting quality in accordance with the standards established in the conduct-of-operations procedures. Specifically, on October 19, 2006, the on-coming control room supervisor relieved the watch during the loss of feedwater transient, instead of waiting for the plant to be in a stable condition, a self-imposed standard documented in Entergy Operations Procedure EN-OP-115, "Conduct of Operations," Revision 2. Although licensee personnel stated that turnover activities were essentially complete at the time, changing the watch at this time caused the at-the-controls reactor operator and other control room personnel to misunderstand who was in charge of the event response and contributed to the at-the-controls operator not placing the mode switch in the SHUTDOWN position, as required by Procedure AOP-0001, "Reactor Scram," Revision 22. The failure to reposition the mode switch resulted in an inadvertent main steam isolation.
This finding was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system, namely the main feedwater system. A Phase 2 estimation was required because this finding resulted in a loss of function of the steam side of the power conversion system as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones." Using the appropriate plant-specific Phase 2 worksheets, this finding was determined to have very low safety significance because the finding only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This finding has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to implement the roles and responsibilities of the senior reactor operators in the main control room as designed.
The licensee entered this performance deficiency into their corrective action program for resolution.
 
Inspection Report# : 2006013 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to identify a degraded condition of steam leak detection system Transmitter E31-N084B A self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified involving the failure to identify a degraded condition affecting the steam leak detection and Division II isolation logic for residual heat removal/reactor core isolation cooling systems. The degraded condition resulted in a spurious isolation of the reactor core isolation cooling system during power operations on November 23, 2006. This issue was entered into the licensees corrective action program as CR-RBS-2006-04460.
The finding was more than minor because it is associated with the mitigating system cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Manual Chapter 0609, "Significance Determination Process," a Phase 2 analysis concluded that the finding was of very low safety significance. The cause of the finding is related to the crosscutting aspect of problem identification and resolution in that the licensee failed to completely and accurately identify the condition that caused a previous isolation of the reactor core isolation cooling system on October 1, 2004. This failure resulted in the spurious reactor core isolation cooling system isolation on November 23, 2006.
Inspection Report# : 2006005 (pdf)
Barrier Integrity Significance:        Jun 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Instructions Resulted in Exceeding Load Line Analysis Limit A self-revealing noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow procedure. Specifically, during control rod withdrawal a reactor engineer noted that reactor power, as calculated by a heat balance, was inconsistent with predicted power. Although this inconsistency was identified the reactor engineers and operators failed to fully evaluate this condition, as required by procedure, and continued with power ascension resulting in an automatic rod withdrawal block. Upon further review the event was caused from feed flow and temperature data not automatically updating resulting in calculated power being less than actual power. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-01691.
The finding was more than minor because it was associated with the barrier integrity cornerstone attribute of configuration control and it affected the cornerstone objective to provide reasonable assurance that physical design barriers, such as fuel cladding, protect the public from radio-nuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have a very low safety significance because it did not have the potential to affect the integrity of the RCS barrier. The cause of this finding is related to the human performance cross cutting component of work practices because neither self nor peer checking actions prevented the automatic rod withdrawal block (H.4(a)).
Inspection Report# : 2007003 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate work instructions result in isolation of annulus pressure control system and automatic start of the Division II standby gas treatment system A self-revealing, noncited violation of Technical Specification 5.4.1.a was identified involving the failure to provide adequate maintenance instructions for replacement of relays in the Division I standby gas treatment system initiation
 
logic. As a result, on November 21, 2006, during relay replacement, the annulus pressure control system tripped and the Division II standby gas treatment system automatically initiated. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04445.
This finding was more than minor because it is associated with the barrier integrity cornerstone attribute of human performance affecting the cornerstone objective to provide reasonable assurance that the secondary containment barrier protects the public from radionuclide releases caused by accidents and events. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because only the standby gas treatment system was affected. The cause of the finding is related to the crosscutting element of human performance in that the licensee failed to provide complete, accurate, and up-to-date instructions in the work package to replace the relays in the Division I standby gas treatment system initiation logic.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow procedure resulted in loss of power to safety-related instrumentation bus and isolation of reactor water cleanup system A self-revealing, noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow Procedure SOP-0048, 120 Vac System, Revision 303. Due to ineffective self- and peer-checking a procedure step was missed, resulting in inadvertent isolation of the reactor water cleanup and the suppression pool cooling and cleanup systems. This issue was entered into the licensee's corrective action program as CR-RBS-2006-03874.
The finding was more than minor because the loss of the reactor water cleanup system, providing reactor water chemistry control, affects the fuel barrier integrity cornerstone attribute of configuration control. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because only the fuel cladding barrier was affected. The cause of the finding is related to the crosscutting element of human performance in that operations personnel failed to make proper use of human performance techniques of self- and peer-checking.
Inspection Report# : 2006005 (pdf)
Significance:        Dec 31, 2006 Identified By: Self-Revealing Item Type: FIN Finding Newly installed reactor water cleanup pump coupling failed because it was beyond its expected service lifetime A self-revealing finding was identified involving the installation of a pump coupling that exceeded vendor shelf- and service-life recommendations on November 15, 2006. As a result, the reactor water cleanup Pump A coupling failed on November 28, 2006, requiring operators to remove from service the reactor water cleanup pump and a demineralizer affecting the primary means of reactor water chemistry control. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04488 and -04517.
The finding is greater than minor because it would become a more significant safety concern if left uncorrected, since failure of similar couplings affecting other plant components, such as the drywell floor and equipment drain pumps, would require a plant shutdown to make repairs. The finding affected the barrier integrity cornerstone. Using the NRC Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because the finding only affected the fuel cladding barrier.
Inspection Report# : 2006005 (pdf)
Emergency Preparedness Occupational Radiation Safety
 
Significance:      Jul 13, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conspicuously Post Radiation Areas The team identified a noncited violation of 10 CFR 20.1902(a) because the licensee failed to post radiation areas in the radwaste building with a conspicuous sign or signs bearing the radiation symbol and the words Caution, Radiation Area. The licensee posted radiation area signs only at the entrances to the different elevations of the building, instead of at the discrete radiation areas, even though most of the radwaste building was not a radiation area.
Dose rates in unposted radiation areas were as high as 15 millirems per hour. As corrective action, the licensee posted the discrete areas. Additional corrective action is still being evaluated.
The finding was greater than minor because it was associated with one of the cornerstone attributes (exposure control and monitoring) and the finding affected the Occupational Radiation Safety cornerstone objective, in that, uninformed workers could unknowingly accrue additional radiation dose. Using the Occupational Radiation Safety Significance Determination Process, the team determined that the finding was of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Also, this finding had a cross-cutting aspect in the area of human performance and component of work control because the licensee did not coordinate work activities by incorporating actions to address the need to keep personnel apprised of plant conditions that may affect work activities.
Inspection Report# : 2007010 (pdf)
Significance:      Dec 31, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Licensee failed to perform a radiological survey in off-gas sample room after radiological conditions had changed A self-revealing, noncited violation of 10 CFR 20.1501(a)(2) was identified involving the failure of radiation protection personnel to perform a survey in the off-gas sample room during main condenser leak testing. As a result, when a chemistry technician entered the room to obtain a grab sample, his electronic alarming dosimeter alarmed unexpectedly. When another chemistry technician reached into the room to perform a survey of the test equipment, his dosimeter also alarmed. It was later determined that they were exposed to a dose rate of 440 and 521 millirem per hour, respectively. This issue was entered into the licensee's corrective action program as CR-RBS-2006-04340.
The finding was more than minor because it was associated with the occupational radiation safety cornerstone attribute of programs and processes, such as the monitoring of radiological conditions, specifically the failure to perform a survey following changes in radiological conditions in the off-gas sample room, and affects the associated cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Utilizing Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for an overexposure, or (4) an impaired ability to assess dose.
The cause of the finding was related to the crosscutting element of problem identification and resolution in that the licensee failed to communicate to affected personnel in a timely manner internal operating experience, specifically, while there was off-gas flow through the condenser leak test equipment, radiological conditions would increase.
Inspection Report# : 2006005 (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 Significance: N/A May 21, 2007 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed approximately 227 condition reports, work orders, engineering evaluations, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities. On most occasions, the team determined that the licensee adequately identified, evaluated, prioritized, and implemented timely and effective corrective actions for conditions adverse to quality. However, the team concluded that the licensee had experienced some continuing challenges in all three areas based upon the number of issues identified during the last 15 months. Examples of poor engineering evaluations continued during this assessment period; however, the licensee had recognized this deficiency and had taken actions to address the weakness. The licensee had also implemented actions to improve their ability to correctly identify and take appropriate actions in response to the Substantive Crosscutting Issue in Problem Identification and Resolution identified in 2006. The licensee improved in their coordination among plant processes when closing condition reports to other corrective action or work control documents although some instances of incorrect closure had recently been identified.
Overall, the licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items into the corrective action program. The licensee appropriately used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective Quality Assurance audits and self-assessments, as demonstrated by self-identification of poor corrective action program performance and identification of ineffective corrective actions. The team concluded that the licensee maintains an appropriate safety conscious work environment. The team concluded from interviews that, although no safety conscious work environment concerns existed, the complaints related to general culture factors that have been stated for the last two safety culture surveys, if not addressed, might result in safety conscious work environment concerns.
Inspection Report# : 2007009 (pdf)
Last modified : December 07, 2007
 
River Bend 1 4Q/2007 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Vendor Recommendations A self-revealing finding was identified involving the failure to implement 1998 vendor recommendations associated with the potential for vibration induced degradation of recirculation loop gate valves. This resulted in the failure to identify and implement timely corrective actions prior to disk to stem separation of recirculation Pump A discharge gate valve that occurred on May 21, 2007. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-02113.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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 the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have a very low safety significance because the finding did not contribute to the likelihood that mitigation equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007003 (pdf)
Significance:      Jun 30, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Work Instructions The inspectors identified a finding involving inadequate maintenance instructions for opening a stuck closed feedwater regulating Valve A isolation valve. Specifically, the instructions failed to account for the system being pressurized resulting in unexpected valve stem movement while technicians were removing the manual operator from the valve on June 10, 2007. This deficiency could have resulted in personnel harm or an unexpected and uncontrolled plant transient. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-02576.
The finding was more than minor because it could become a more significant safety concern if left uncorrected. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have very low safety significance because the deficiency did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. No violation of NRC requirements occurred. The cause of this finding was related to the human performance crosscutting component of resources because the licensee did not ensure a complete and accurate work package was available prior to the start of the job (H.2(c)).
Inspection Report# : 2007003 (pdf)
Significance:      Jun 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Install Scram Discharge Instrument Volume Vent Plug A self-revealing noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow a surveillance procedure for scram discharge instrument volume water level channel calibration. Specifically, on February 9, 2007, an instrument line plug was not replaced following surveillance testing. As a result, on May 5, 2007, following a reactor scram, reactor water sprayed out of the scram discharge instrument volume and
 
contaminated some accessible portions of the containment building causing three inadvertent personnel contamination events. This issue was entered into the licensees corrective action program as condition Report CR-RBS-2007-01809.
The finding was more than minor because it was associated with the initiating event cornerstone attribute of equipment performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency resulted in a reactor coolant leak greater than the Technical Specification limit for identified reactor coolant system leakage. Using the plant-specific Phase 2 risk-informed notebook, this violation was determined to have very low safety significance because the violation only increased the likelihood of a small-break loss of coolant accident by a very small amount and mitigation capability was unaffected. The cause of the finding was related to the human performance crosscutting component of work practices because neither self nor peer checking actions identified the failure to replace the vent plug (H.4(a)).
Inspection Report# : 2007003 (pdf)
Significance:        Mar 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Maintenance Instructions A self-revealing finding was identified involving the failure of maintenance personnel to follow maintenance instructions resulting in the failure to properly seal the desiccant retention strainer of an instrument air dryer. As a result, desiccant was released from the dryer tower and became lodged in an outlet shuttle valve causing it to stick open that resulted in lowering the instrument air header pressure. This condition caused operators to enter the abnormal operating procedure for loss of instrument air, an automatic start of standby air compressors, and the automatic cross-connect of service air to the instrument air header. These actions restored instrument air pressure preventing a significant plant transient. This issue was entered into the licensee's corrective action program as CR-RBS-2007-00438.
The finding was more than minor because it would become a more significant safety concern if left uncorrected in that an air dryer failure could result in a complete loss of instrument air. The finding affected the initiating event cornerstone. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because there was no actual loss of instrument air. The cause of the finding was related to the crosscutting element of human performance in that the maintenance technicians failed to properly self and peer check the adequacy of the retention strainer seal during maintenance of instrument air Dryer 2 on January 12, 2007. As a result, desiccant was released causing an outlet shuttle valve to stick open.
Inspection Report# : 2007002 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to identify and correct discrepancies between the design function and observed response of the feedwater isolation valves prior to reactor restart An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to identify and correct a condition adverse to quality in a timely manner.
Specifically, following the reactor scram on October 19, 2006, licensee personnel failed to properly evaluate discrepancies between the expected response of Feedwater Isolation Valves FWS-MOV7A and FWS-MOV7B, operator observation of valve indication, and indication of actual plant parameters affected by the valves, prior to restarting the reactor on October 22, 2006.
This violation was greater than minor because it was associated with the problem identification and resolution and the human performance attributes of the initiating events cornerstone and affected 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. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers
 
Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the violation only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program that ensured timely resolution of conditions adverse to quality.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of reactor operators to perform an adequate control board walkdown resulting in failure to identify that feedwater isolation valves were closing A self-revealing noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed conduct-of-operations procedures. Specifically, on October 19, 2006, two senior reactor operators (one on-coming and one off-going),
conducting turnover activities, and the at-the-controls reactor operator failed to identify that the push buttons for Main Feedwater Isolation Valves 7A and 7B were out of alignment upon panel inspection during panel walk downs conducted in accordance with Entergy Operations Procedure EN-OP-115, "Conduct of Operations," Revision 2.
This violation was greater than minor because it was associated with the human performance attribute of the initiating events cornerstone and affected 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. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was initially determined to have very low safety significance because the violation only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to effectively use human error prevention techniques, such as self and peer checking.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to identify, place in the corrective action program, and correct deficiencies with Chart Recorder C33-R608 prior to restarting the reactor An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to identify and correct a condition adverse to quality in a timely manner.
Specifically, on October 19, 2006, a licensed reactor operator noted a nonconforming condition with Strip Chart Recorder C33-R608 following the fall of the chart paper mechanism and discussed this with his supervision.
However, this condition was not documented in the condition reporting process, the recorder was not properly inspected and repaired by qualified maintenance technicians prior to reactor restart, and at least one member of the on-site safety review committee may have been misinformed about the extent and composition of the evaluation and repair activities conducted on control room recorders prior to authorizing plant restart on October 22, 2006.
This finding was greater than minor because it was associated with the problem identification and resolution and the human performance attributes 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 because the chart recorder was left in a condition that had resulted in a reactor scram. A Phase 2 estimation
 
was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this finding was determined to be of very low safety significance because it only impacted the plant for a 2-day period. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program with a low threshold for identifying issues.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Licensee personnel failed to provide complete corrective actions to address the probable cause of the October 19, 2006, scram, prior to restarting the reactor An NRC-identified noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to correct a condition adverse to quality. Specifically, following the reactor scram on October 19, 2006, licensee personnel determined that the probable cause of the scram was a human performance error while handling the chart recorder. However, while significant corrective actions were taken, these actions did not completely address this probable cause prior to restarting the reactor on October 22, 2006, in that, expectations for working over control panels were not fully conveyed.
This violation was greater than minor because it was associated with the problem identification and resolution and the human performance attributes 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 because expectations and/or guidance were not provided to licensed operators on how to correct paper take up problems on strip chart recorders while minimizing the risk of dropping components on controls. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency represented an increase in the likelihood of both a reactor trip and the likelihood that the power conversion system would be unavailable. Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to be of very low safety significance because it only impacted the plant for a limited period of time. This finding has a cross-cutting aspect in the area of problem identification and resolution, in that, the licensee did not implement a corrective action program that ensured timely resolution of conditions adverse to quality.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Mitigating Systems Significance:        Sep 29, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Removing Control Building Chilled Water System from Service An NRC identified noncited violation of 10 CFR 50.65 (a)(4) was identified for the failure to assess and manage the increase in risk that may result from proposed maintenance activities on the control building chilled water system.
This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2007-03059.
Using NRC Manual Chapter 0612, Appendix B, Section 3, Item 5(h), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions that changed the outcome of the assessment. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk
 
Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time periods is less than 1.0E-6 Inspection Report# : 2007004 (pdf)
Significance:      Apr 27, 2007 Identified By: NRC Item Type: FIN Finding Foreign Material Found in Residual Heat Removal Room Sump Pump Discharge Check Valve The team identified a finding because the licensee failed to address control of foreign material in the Train B residual heat removal room in June 2003. Consequently, on March 5, 2007, maintenance technicians found foreign material in one of the sump pump discharge check valves. This failure to control foreign material resulted in sump high level alarms, which had caused the operators to enter the emergency operating procedure for auxiliary building room flooding on three different occasions. The licensee documented this deficiency in Condition Report 2007-00859.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability of the residual heat removal system. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because there was no actual loss of the residual heat removal system function and it did not screen as potentially risk significant for an internal flooding event. The cause of the finding was related to the crosscutting element of human performance work practices in that licensee management failed to communicate and enforce compliance with the site foreign material control program.
Inspection Report# : 2007009 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Implement Freeze Protection Compensatory Measures The inspectors identified a finding involving the failure of operators to implement compensatory measures for cold weather conditions when a ventilation heater for a safety related standby cooling tower pipe chase was out of service during the winters from 2003 through 2006. This issue was entered into the licensees corrective action program as CR-RBS-2007-00399.
The finding was more than minor because it was associated with the mitigating system cornerstone attribute of equipment performance and affected the associated cornerstone objective. The finding was determined to have very low safety significance because it did not result in a actual loss of the standby service water system and it was determined by a Phase 3 analysis not to be risk significant due to external events. The cause of the finding was related to the crosscutting aspect of problem identification and resolution in that the licensee failed to identify that freeze protection equipment in the area was out of service each winter from 2003 through 2006 requiring compensatory measures.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct a Degraded Residual Heat Removal System Valve A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified involving the failure to promptly identify and correct a condition adverse to quality. Specifically, on August 3, 2005, residual heat removal Train A fuel pool cooling assist Valve E12-MOVF037A failed to fully close during actuation. The failure to correct the problem resulted in recurrence of the valve failing to fully close on April 11, 2006, and January 7, 2007. This issue was entered into the licensee's corrective action program as CR-RBS-2006-01326.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and affected the associated cornerstone objective. The finding had very low safety significance because it did not represent a loss of the residual heat removal system safety function. The cause of the
 
finding was related to the crosscutting element of problem identification and resolution in that the licensee did not thoroughly evaluate the problem such that the resolution would address the cause of the failure of Valve E12-MOVF037A to fully close on August 3, 2005.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Instructions for Installation of a Compression Fitting The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings, for the licensees failure to provide adequate work instructions for repairing a failed tubing compression fitting on the Division I emergency diesel generator jacket cooling water system. Specifically, the repair inappropriately had tubing entering a compression fitting at an angle that could result in failure as had previously been encountered on the same fitting. This issue was entered into the licensee's corrective action program as CR-RBS-2007-01496.
The finding was more than minor because it would become a more significant event if left uncorrected in that failure to install and repair tubing fittings correctly can lead to subsequent failure. The finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because the repair did not result in an actual loss of function of the Division I emergency diesel generator. The cause of the finding was related to the crosscutting element of human performance in that the licensee did not effectively communicate expectations for proper assembly of tubing fittings on safety related equipment.
Inspection Report# : 2007002 (pdf)
Significance:      Feb 28, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to place the reactor mode switch in the SHUTDOWN position following a reactor scram as required by abnormal operating procedures A self-revealing noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures. Specifically, the at-the-controls operator failed to perform an immediate action required by Abnormal Operating Procedure AOP-0001, "Reactor Scram," Revision 22, which required him to place the mode switch in the SHUTDOWN position. The failure to reposition the mode switch resulted in an inadvertent main steam isolation, complicating the scram recovery.
This violation was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system. A Phase 2 estimation was required because this violation represented a loss of function of the steam side of the power conversion system as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones."
Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the errors only impacted the plant for a short period of time and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to effectively use human error prevention techniques.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:      Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify that the reactor mode switch was in the SHUTDOWN position following a reactor scram as required by emergency operating procedures
 
An NRC-identified noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures.
Specifically, the control room supervisor failed to follow Emergency Operating Procedure EOP-0001, "Reactor Pressure Vessel Control," Revision 20, which required him to verify that the mode switch was in the SHUTDOWN position. The failure to reposition the mode switch resulted in an inadvertent main steam isolation, complicating the scram recovery.
This violation was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system. A Phase 2 estimation was required because this violation represented a loss of function of the steam side of the power conversion system, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones."
Using the appropriate plant-specific Phase 2 worksheets, this violation was determined to have very low safety significance because the errors only impacted the plant for a short period of time and the power conversion system was actually recoverable. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to provide adequate management oversight in this situation.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to permit the safety/relief valves to cycle in automatic and to manually operate the safety/relief valves without driving level outside the prescribed level band as required by AOPs An NRC-identified noncited violation of Technical Specification, Section 5.4, "Procedures," was identified for the failure of licensee personnel to accomplish activities affecting quality in accordance with prescribed procedures.
Specifically, licensed operators operated the safety/relief valves manually contrary to Abnormal Operating Procedure AOP-0001, OSP-0053, Attachment 1B, "Post Scram Pressure Control Strategies," Revision 5, requirements to operate them in automatic with the main steam isolation valves closed. Additionally, operators failed to manually operate the safety/relief valves, as required, to control pressure in the prescribed pressure band, without driving level outside the prescribed level band.
This violation was more than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system because manual actions affect licensed operator capability to perform simultaneous actions. Using the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," the finding was of very low safety significance because it did not represent a loss of safety function nor did it screen as potentially significant to external initiators. This violation has a cross-cutting aspect in the area of human performance, work practices component associated with the effectiveness of communicating expectations regarding procedural compliance.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Significance:        Feb 28, 2007 Identified By: NRC Item Type: FIN Finding Senior reactor operator relieved the watch during a transient without waiting for the plant to be in a stable condition, resulting in an inadvertent main steam isolation The team identified a finding for the failure of licensed operators to accomplish activities affecting quality in accordance with the standards established in the conduct-of-operations procedures. Specifically, on October 19, 2006, the on-coming control room supervisor relieved the watch during the loss of feedwater transient, instead of waiting for the plant to be in a stable condition, a self-imposed standard documented in Entergy Operations Procedure EN-OP-115, "Conduct of Operations," Revision 2. Although licensee personnel stated that turnover activities were essentially complete at the time, changing the watch at this time caused the at-the-controls reactor operator and other control
 
room personnel to misunderstand who was in charge of the event response and contributed to the at-the-controls operator not placing the mode switch in the SHUTDOWN position, as required by Procedure AOP-0001, "Reactor Scram," Revision 22. The failure to reposition the mode switch resulted in an inadvertent main steam isolation.
This finding was greater than minor because it was associated with the human performance attribute and affected the mitigating systems cornerstone objective to ensure the availability, reliability, or function of a system or train in a mitigating system, namely the main feedwater system. A Phase 2 estimation was required because this finding resulted in a loss of function of the steam side of the power conversion system as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones." Using the appropriate plant-specific Phase 2 worksheets, this finding was determined to have very low safety significance because the finding only increased the initiating event likelihood by a very small amount and the power conversion system was actually recoverable. This finding has a cross-cutting aspect in the area of human performance, work practices component associated with the failure to implement the roles and responsibilities of the senior reactor operators in the main control room as designed.
The licensee entered this performance deficiency into their corrective action program for resolution.
Inspection Report# : 2006013 (pdf)
Barrier Integrity Significance:        Jun 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Instructions Resulted in Exceeding Load Line Analysis Limit A self-revealing noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow procedure. Specifically, during control rod withdrawal a reactor engineer noted that reactor power, as calculated by a heat balance, was inconsistent with predicted power. Although this inconsistency was identified the reactor engineers and operators failed to fully evaluate this condition, as required by procedure, and continued with power ascension resulting in an automatic rod withdrawal block. Upon further review the event was caused from feed flow and temperature data not automatically updating resulting in calculated power being less than actual power. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-01691.
The finding was more than minor because it was associated with the barrier integrity cornerstone attribute of configuration control and it affected the cornerstone objective to provide reasonable assurance that physical design barriers, such as fuel cladding, protect the public from radio-nuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have a very low safety significance because it did not have the potential to affect the integrity of the RCS barrier. The cause of this finding is related to the human performance cross cutting component of work practices because neither self nor peer checking actions prevented the automatic rod withdrawal block (H.4(a)).
Inspection Report# : 2007003 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Sep 29, 2007 Identified By: NRC Item Type: NCV NonCited Violation
 
Failure to Survey Following Containment Atmosphere Radiation Monitor Particulate Channel Alarms An NRC-identified noncited violation of 10 CFR 20.1501(a) was identified involving multiple failures to perform radiological surveys to evaluate radiological hazards following control room alarms of the Containment Atmosphere Radiation monitor particulate channel. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2007-04415.
This finding is more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of program and process, and affects the cornerstone objective to ensure the adequate protection of a workers health and safety from exposure to radiation because it could have resulted in workers being exposed to higher radiation levels. When processed through the Occupational Radiation Safety Significance Determination Process, the finding is determined to be of very low safety significance because it is not an as low as is reasonably achievable finding, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
The finding has a crosscutting aspect in the area of human performance, specifically the work control component, because the licensee failed to appropriately coordinate work activities by incorporating actions to address the impact of the work on different job activities and the need for work groups to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance (H.3(b)).
Inspection Report# : 2007004 (pdf)
Significance:      Jul 13, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conspicuously Post Radiation Areas The team identified a noncited violation of 10 CFR 20.1902(a) because the licensee failed to post radiation areas in the radwaste building with a conspicuous sign or signs bearing the radiation symbol and the words Caution, Radiation Area. The licensee posted radiation area signs only at the entrances to the different elevations of the building, instead of at the discrete radiation areas, even though most of the radwaste building was not a radiation area.
Dose rates in unposted radiation areas were as high as 15 millirems per hour. As corrective action, the licensee posted the discrete areas. Additional corrective action is still being evaluated.
The finding was greater than minor because it was associated with one of the cornerstone attributes (exposure control and monitoring) and the finding affected the Occupational Radiation Safety cornerstone objective, in that, uninformed workers could unknowingly accrue additional radiation dose. Using the Occupational Radiation Safety Significance Determination Process, the team determined that the finding was of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Also, this finding had a cross-cutting aspect in the area of human performance and component of work control because the licensee did not coordinate work activities by incorporating actions to address the need to keep personnel apprised of plant conditions that may affect work activities.
Inspection Report# : 2007010 (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
 
Significance: N/A May 21, 2007 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed approximately 227 condition reports, work orders, engineering evaluations, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities. On most occasions, the team determined that the licensee adequately identified, evaluated, prioritized, and implemented timely and effective corrective actions for conditions adverse to quality. However, the team concluded that the licensee had experienced some continuing challenges in all three areas based upon the number of issues identified during the last 15 months. Examples of poor engineering evaluations continued during this assessment period; however, the licensee had recognized this deficiency and had taken actions to address the weakness. The licensee had also implemented actions to improve their ability to correctly identify and take appropriate actions in response to the Substantive Crosscutting Issue in Problem Identification and Resolution identified in 2006. The licensee improved in their coordination among plant processes when closing condition reports to other corrective action or work control documents although some instances of incorrect closure had recently been identified.
Overall, the licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items into the corrective action program. The licensee appropriately used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective Quality Assurance audits and self-assessments, as demonstrated by self-identification of poor corrective action program performance and identification of ineffective corrective actions. The team concluded that the licensee maintains an appropriate safety conscious work environment. The team concluded from interviews that, although no safety conscious work environment concerns existed, the complaints related to general culture factors that have been stated for the last two safety culture surveys, if not addressed, might result in safety conscious work environment concerns.
Inspection Report# : 2007009 (pdf)
Last modified : February 04, 2008
 
River Bend 1 1Q/2008 Plant Inspection Findings Initiating Events Significance:        Mar 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Internal Operating Experience Not Used to Prevent Recurrence of Reactor Recirculation FCV Runbacks The inspectors identified a noncited violation of Technical Specification 5.4.1.a for an inadequate procedure for securing a reactor feedwater pump. Specifically, the licensee failed to incorporate internal operating experience into the procedure. As a result, a reactor recirculation flow control valve runback resulting from a known reactor vessel water level loop tolerance issue recurred, resulting in an unplanned power reduction. This issue was entered into the licensees corrective action program as Condition Report RBS-2007-4749.
The finding is more than minor since it affects the human performance area 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. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding has very low safety significance since it did not contribute to both the likelihood of a reactor scram and the likelihood that mitigating equipment would not have been available.
Inspection Report# : 2008002 (pdf)
Significance:        Mar 29, 2008 Identified By: Self-Revealing Item Type: FIN Finding Condensate Demineralizer Tank Liner Failure A self-revealing finding was identified for the failure to properly repair condensate Demineralizer 1E tank liner prior to returning it to service. As a result, failure of the liner resulted in approximately 20,000 gallons of radiological contaminated condensate being spilled from the manway flange. Operations lowered reactor power from 90 percent to 82 percent to conserve condensate system inventory. This issue was entered into the licensees corrective action program as Condition Report RBS-2007-5440.
The finding is greater than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was considered to be a transient initiator contributor which contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available and, therefore, screened to Phase 2. Using the Phase 2 worksheets, the inspectors assumed that successful recovery of the condensate system from the leak was highly likely and determined the finding to be of very low safety significance. This finding has crosscutting aspects associated with human performance in the area of resources in that a complete, accurate, and up-to-date work package was not available to assure nuclear safety
[H.2(c)].
Inspection Report# : 2008002 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation RPS Terminal Board Loose Connection Results in a Reactor Scram A self-revealing noncited violation of 10 CFR Part 50 Appendix B, Criterion V was identified involving the failure to adequately torque reactor protection system electrical terminal board connections during initial construction. This
 
failure resulted in a loose terminal connection causing thermal degradation that subsequently resulted in an automatic reactor scram during average power range monitor surveillance testing. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2007-04264.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and that mitigating equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Preventive Maintenance Strategy Results in a Breaker's Failure to Promptly Open Due to Hardened Grease Results in a Complicated Reactor Scram A self-revealing Green noncited violation of 10 CFR 50.65(A)(3) was identified for failure to incorporate internal and external operating experience into preventive maintenance activities to prevent industry known electrical circuit breaker deficiencies. Specifically, inadequate breaker maintenance, leading to grease hardening degradation, resulted in inadequate electrical fault protection on November 7, 2007. The failure to adequately isolate the electrical fault resulted in a complicated reactor scram involving the loss of the main condenser and reactor feedwater. The licensee entered this into their corrective action program as CR-RBS-2007-04922.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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. The inspectors evaluated the finding using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The finding required a Phase 2 analysis because the finding contributed to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. A senior reactor analyst estimated the risk of the subject finding using the Risk-Informed Inspection Notebook for River Bend Station, Unit 1, Revision 2.1a. The analyst determined the finding was of very low safety significance.
This finding has a crosscutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Preventative Maintenance Results in a Plant Down Power A self-revealing finding was identified for failure to perform adequate preventive maintenance for control panels associated with providing make up water to the circulating water system. Adequate preventative maintenance was not performed on this system, resulting in failure, based on an inappropriate run to failure classification of this equipment.
The failure of this system resulted in a significant unplanned reduction in reactor power to 20 percent. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2007-04447.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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, Significance Determination Process, Phase 1 Worksheet, the finding has very low safety significance (Green) since the finding did not contribute to both the likelihood of a reactor trip and that mitigating equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007005 (pdf)
 
Significance:      Jun 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Vendor Recommendations A self-revealing finding was identified involving the failure to implement 1998 vendor recommendations associated with the potential for vibration induced degradation of recirculation loop gate valves. This resulted in the failure to identify and implement timely corrective actions prior to disk to stem separation of recirculation Pump A discharge gate valve that occurred on May 21, 2007. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-02113.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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 the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have a very low safety significance because the finding did not contribute to the likelihood that mitigation equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007003 (pdf)
Significance:      Jun 30, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Work Instructions The inspectors identified a finding involving inadequate maintenance instructions for opening a stuck closed feedwater regulating Valve A isolation valve. Specifically, the instructions failed to account for the system being pressurized resulting in unexpected valve stem movement while technicians were removing the manual operator from the valve on June 10, 2007. This deficiency could have resulted in personnel harm or an unexpected and uncontrolled plant transient. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-02576.
The finding was more than minor because it could become a more significant safety concern if left uncorrected. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have very low safety significance because the deficiency did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. No violation of NRC requirements occurred. The cause of this finding was related to the human performance crosscutting component of resources because the licensee did not ensure a complete and accurate work package was available prior to the start of the job (H.2(c)).
Inspection Report# : 2007003 (pdf)
Significance:      Jun 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Install Scram Discharge Instrument Volume Vent Plug A self-revealing noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow a surveillance procedure for scram discharge instrument volume water level channel calibration. Specifically, on February 9, 2007, an instrument line plug was not replaced following surveillance testing. As a result, on May 5, 2007, following a reactor scram, reactor water sprayed out of the scram discharge instrument volume and contaminated some accessible portions of the containment building causing three inadvertent personnel contamination events. This issue was entered into the licensees corrective action program as condition Report CR-RBS-2007-01809.
The finding was more than minor because it was associated with the initiating event cornerstone attribute of equipment performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. A Phase 2 estimation was required, as determined by the Manual Chapter 0609, Appendix A, Phase 1 Worksheet, "SDP Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones," because the associated performance deficiency resulted in a reactor coolant leak greater than the Technical Specification limit for identified reactor coolant system
 
leakage. Using the plant-specific Phase 2 risk-informed notebook, this violation was determined to have very low safety significance because the violation only increased the likelihood of a small-break loss of coolant accident by a very small amount and mitigation capability was unaffected. The cause of the finding was related to the human performance crosscutting component of work practices because neither self nor peer checking actions identified the failure to replace the vent plug (H.4(a)).
Inspection Report# : 2007003 (pdf)
Mitigating Systems Significance:        Mar 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Improper Design Control for Evaluating Emergency Diesel Generator Turbocharger Combustion Air Pipe Stresses The inspectors identified a noncited violation of Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control,"
for failure to incorporate accurate design information into a calculation to determine emergency diesel generator turbocharger discharge combustion air pipe stresses. This resulted in pipe failure. Specifically, a calculation assumed nonconservative pipe wall thicknesses and process air temperatures, treated pipe end points as rigid anchors and failed to use stress intensification factors. This resulted in low calculated pipe stresses. With appropriately calculated pipe stress values, Entergy personnel could reasonably have been expected to adequately modify the combustion air piping to preclude subsequent failures. This issue was entered into the licensees corrective action program as Condition Report RBS-2008-2869.
This issue was determined to be more than minor because it affected the mitigating systems cornerstone objective and was similar to Manual Chapter 0612, Appendix E, Example 3.j because the errors were considered more than a minor calculation error in that the deficiency failed to identify the high pipe wall stresses that significantly reduced the overall allowable material strength margin. Later pipe and weld flaws developed at the intercooler adapter and turbocharger end connections that rendered the emergency diesel generator Division 2 inoperable. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that the issue was of very low safety significance (Green) because it did not screen as risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008002 (pdf)
Significance:        Sep 29, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Removing Control Building Chilled Water System from Service An NRC identified noncited violation of 10 CFR 50.65 (a)(4) was identified for the failure to assess and manage the increase in risk that may result from proposed maintenance activities on the control building chilled water system.
This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2007-03059.
Using NRC Manual Chapter 0612, Appendix B, Section 3, Item 5(h), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions that changed the outcome of the assessment. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time periods is less than 1.0E-6 Inspection Report# : 2007004 (pdf)
Significance:        Apr 27, 2007 Identified By: NRC Item Type: FIN Finding Foreign Material Found in Residual Heat Removal Room Sump Pump Discharge Check Valve The team identified a finding because the licensee failed to address control of foreign material in the Train B residual
 
heat removal room in June 2003. Consequently, on March 5, 2007, maintenance technicians found foreign material in one of the sump pump discharge check valves. This failure to control foreign material resulted in sump high level alarms, which had caused the operators to enter the emergency operating procedure for auxiliary building room flooding on three different occasions. The licensee documented this deficiency in Condition Report 2007-00859.
The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and affected the associated cornerstone objective to ensure the availability of the residual heat removal system. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance because there was no actual loss of the residual heat removal system function and it did not screen as potentially risk significant for an internal flooding event. The cause of the finding was related to the crosscutting element of human performance work practices in that licensee management failed to communicate and enforce compliance with the site foreign material control program.
Inspection Report# : 2007009 (pdf)
Barrier Integrity Significance:      Mar 29, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Reactor Startup Procedure Results in Six Control Rod Withdrawal Errors A self-revealing noncited violation of Technical Specification 5.4.1.a occurred when River Bend Station reactor operators failed to comply with General Operating Procedure GOP 000-1, Plant Start Up. Specifically operators withdrew six control rods two notches past the target out notch position specified in Reactivity Control Plan RCP 03. No fuel damage resulted from these errors. This issue was entered into the licensees corrective action program as Condition Report RBS-2008-2174.
This finding was more than minor because the finding affected the barrier integrity cornerstone attributes of configuration control and human performance and adversely impacts the cornerstones objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radio nuclide releases caused by accidents or events. The inspectors completed a Phase 1 significance determination using Manual Chapter 0609 Appendix A, Significance Determination Process Phase 1 screening worksheet, and determined the finding to be of very low safety significance (Green) because the performance issue only degraded the fuel cladding barrier. This finding had crosscutting aspects associated with human performance in the area of work practices in that the reactor operators failed to use self-check and peer-check during control rod reactivity manipulations (H.4.a).
Inspection Report# : 2008002 (pdf)
Significance:      Jun 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Instructions Resulted in Exceeding Load Line Analysis Limit A self-revealing noncited violation of Technical Specification 5.4.1.a was identified involving the failure to follow procedure. Specifically, during control rod withdrawal a reactor engineer noted that reactor power, as calculated by a heat balance, was inconsistent with predicted power. Although this inconsistency was identified the reactor engineers and operators failed to fully evaluate this condition, as required by procedure, and continued with power ascension resulting in an automatic rod withdrawal block. Upon further review the event was caused from feed flow and temperature data not automatically updating resulting in calculated power being less than actual power. This issue was entered into the licensee's corrective action program as condition Report CR-RBS-2007-01691.
The finding was more than minor because it was associated with the barrier integrity cornerstone attribute of configuration control and it affected the cornerstone objective to provide reasonable assurance that physical design barriers, such as fuel cladding, protect the public from radio-nuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have a very low safety significance because it did not have the potential to affect the integrity of the RCS barrier. The cause
 
of this finding is related to the human performance cross cutting component of work practices because neither self nor peer checking actions prevented the automatic rod withdrawal block (H.4(a)).
Inspection Report# : 2007003 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Evaluate the Magnitude and Extent of Radiological Hazards Results in Personnel Contaminations A self-revealing noncited violation of 10 CFR 20.1501(a) was identified for failure to evaluate the magnitude and extent of radiological hazards associated with performing inspections of equipment in the containment building after a reactor trip on May 4, 2007. This failure resulted in six personnel contaminations and uptakes. Followup surveys identified contamination levels of 60 mRad/smear beta/gamma and up to 1300 dpm alpha. Air sample results determined a derived air concentration value of 44 for noble gas. The licensee has placed this event in the radiation protection continuing training program and entered it into their corrective action program as Condition Report CR-RBS-2007-1822.
This 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 in that the failure to evaluate the magnitude and extent of radiological hazards could cause unintentional dose to radiation workers. This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work control because the licensee did not communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit and Radiation Worker Expectations A self-revealing noncited violation of Technical Specification 5.4.1 was identified for failure to follow radiation work permit instructions resulting in a worker entering a posted high radiation area without authorization. On April 20, 2007, an individual received an electronic alarming dosimeter dose rate alarm after entering a posted high radiation area. The individual was signed on to a radiation work permit that did not allow entry into a high radiation area. This violation was entered into licensees corrective action program as Condition Report CR-RBS-2007-1584.
This finding was greater than minor because it was associated with the occupational radiation safety cornerstone attribute of human performance and affected the cornerstone objective in that the failure to follow radiation work permit requirements could cause unintentional dose. This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work practices because the individual involved did not use proper self-checking and entered an area he was not authorized to enter.
Inspection Report# : 2007005 (pdf)
 
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post a Radiation Area An NRC-identified noncited violation of 10 CFR 20.1902(a) was identified for failure to conspicuously post a radiation area. Specifically, the inspector identified an entrance to a radiation area on the 90-foot elevation of the radwaste building that was accessible by a permanently installed ladder from the 65-foot elevation, which was not conspicuously posted as a radiation area. General area dose rates in the area were as high as 7 mrem/hour. This violation was entered into the licensees corrective action program as Condition Report CR-RBS-2007-4954.
This 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 in that the failure to post radiation areas could cause unintentional dose to radiation workers. This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work practices because radiation protection personnel did not adhere to management expectations regarding procedural compliance and following station procedures.
Inspection Report# : 2007005 (pdf)
Significance:      Sep 29, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Survey Following Containment Atmosphere Radiation Monitor Particulate Channel Alarms An NRC-identified noncited violation of 10 CFR 20.1501(a) was identified involving multiple failures to perform radiological surveys to evaluate radiological hazards following control room alarms of the Containment Atmosphere Radiation monitor particulate channel. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2007-04415.
This finding is more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of program and process, and affects the cornerstone objective to ensure the adequate protection of a workers health and safety from exposure to radiation because it could have resulted in workers being exposed to higher radiation levels. When processed through the Occupational Radiation Safety Significance Determination Process, the finding is determined to be of very low safety significance because it is not an as low as is reasonably achievable finding, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
The finding has a crosscutting aspect in the area of human performance, specifically the work control component, because the licensee failed to appropriately coordinate work activities by incorporating actions to address the impact of the work on different job activities and the need for work groups to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance (H.3(b)).
Inspection Report# : 2007004 (pdf)
Significance:      Jul 13, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conspicuously Post Radiation Areas The team identified a noncited violation of 10 CFR 20.1902(a) because the licensee failed to post radiation areas in the radwaste building with a conspicuous sign or signs bearing the radiation symbol and the words Caution, Radiation Area. The licensee posted radiation area signs only at the entrances to the different elevations of the building, instead of at the discrete radiation areas, even though most of the radwaste building was not a radiation area.
Dose rates in unposted radiation areas were as high as 15 millirems per hour. As corrective action, the licensee posted the discrete areas. Additional corrective action is still being evaluated.
The finding was greater than minor because it was associated with one of the cornerstone attributes (exposure control and monitoring) and the finding affected the Occupational Radiation Safety cornerstone objective, in that, uninformed workers could unknowingly accrue additional radiation dose. Using the Occupational Radiation Safety Significance
 
Determination Process, the team determined that the finding was of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Also, this finding had a cross-cutting aspect in the area of human performance and component of work control because the licensee did not coordinate work activities by incorporating actions to address the need to keep personnel apprised of plant conditions that may affect work activities.
Inspection Report# : 2007010 (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 Significance: N/A May 21, 2007 Identified By: NRC Item Type: FIN Finding Identification and Resolution of Problems The team reviewed approximately 227 condition reports, work orders, engineering evaluations, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities. On most occasions, the team determined that the licensee adequately identified, evaluated, prioritized, and implemented timely and effective corrective actions for conditions adverse to quality. However, the team concluded that the licensee had experienced some continuing challenges in all three areas based upon the number of issues identified during the last 15 months. Examples of poor engineering evaluations continued during this assessment period; however, the licensee had recognized this deficiency and had taken actions to address the weakness. The licensee had also implemented actions to improve their ability to correctly identify and take appropriate actions in response to the Substantive Crosscutting Issue in Problem Identification and Resolution identified in 2006. The licensee improved in their coordination among plant processes when closing condition reports to other corrective action or work control documents although some instances of incorrect closure had recently been identified.
Overall, the licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items into the corrective action program. The licensee appropriately used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective Quality Assurance audits and self-assessments, as demonstrated by self-identification of poor corrective action program performance and identification of ineffective corrective actions. The team concluded that the licensee maintains an appropriate safety conscious work environment. The team concluded from interviews that, although no safety conscious work environment concerns existed, the complaints related to general culture factors that have been stated for the last two safety culture surveys, if not addressed, might result in safety conscious work environment concerns.
Inspection Report# : 2007009 (pdf)
Last modified : June 05, 2008
 
River Bend 1 2Q/2008 Plant Inspection Findings Initiating Events Significance:        Mar 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Internal Operating Experience Not Used to Prevent Recurrence of Reactor Recirculation FCV Runbacks The inspectors identified a noncited violation of Technical Specification 5.4.1.a for an inadequate procedure for securing a reactor feedwater pump. Specifically, the licensee failed to incorporate internal operating experience into the procedure. As a result, a reactor recirculation flow control valve runback resulting from a known reactor vessel water level loop tolerance issue recurred, resulting in an unplanned power reduction. This issue was entered into the licensees corrective action program as Condition Report RBS-2007-4749.
The finding is more than minor since it affects the human performance area 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. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding has very low safety significance since it did not contribute to both the likelihood of a reactor scram and the likelihood that mitigating equipment would not have been available.
Inspection Report# : 2008002 (pdf)
Significance:        Mar 29, 2008 Identified By: Self-Revealing Item Type: FIN Finding Condensate Demineralizer Tank Liner Failure A self-revealing finding was identified for the failure to properly repair condensate Demineralizer 1E tank liner prior to returning it to service. As a result, failure of the liner resulted in approximately 20,000 gallons of radiological contaminated condensate being spilled from the manway flange. Operations lowered reactor power from 90 percent to 82 percent to conserve condensate system inventory. This issue was entered into the licensees corrective action program as Condition Report RBS-2007-5440.
The finding is greater than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was considered to be a transient initiator contributor which contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available and, therefore, screened to Phase 2. Using the Phase 2 worksheets, the inspectors assumed that successful recovery of the condensate system from the leak was highly likely and determined the finding to be of very low safety significance. This finding has crosscutting aspects associated with human performance in the area of resources in that a complete, accurate, and up-to-date work package was not available to assure nuclear safety
[H.2(c)].
Inspection Report# : 2008002 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation RPS Terminal Board Loose Connection Results in a Reactor Scram A self-revealing noncited violation of 10 CFR Part 50 Appendix B, Criterion V was identified involving the failure to adequately torque reactor protection system electrical terminal board connections during initial construction. This
 
failure resulted in a loose terminal connection causing thermal degradation that subsequently resulted in an automatic reactor scram during average power range monitor surveillance testing. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2007-04264.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and that mitigating equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Preventive Maintenance Strategy Results in a Breaker's Failure to Promptly Open Due to Hardened Grease Results in a Complicated Reactor Scram A self-revealing Green noncited violation of 10 CFR 50.65(A)(3) was identified for failure to incorporate internal and external operating experience into preventive maintenance activities to prevent industry known electrical circuit breaker deficiencies. Specifically, inadequate breaker maintenance, leading to grease hardening degradation, resulted in inadequate electrical fault protection on November 7, 2007. The failure to adequately isolate the electrical fault resulted in a complicated reactor scram involving the loss of the main condenser and reactor feedwater. The licensee entered this into their corrective action program as CR-RBS-2007-04922.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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. The inspectors evaluated the finding using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The finding required a Phase 2 analysis because the finding contributed to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. A senior reactor analyst estimated the risk of the subject finding using the Risk-Informed Inspection Notebook for River Bend Station, Unit 1, Revision 2.1a. The analyst determined the finding was of very low safety significance.
This finding has a crosscutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Preventative Maintenance Results in a Plant Down Power A self-revealing finding was identified for failure to perform adequate preventive maintenance for control panels associated with providing make up water to the circulating water system. Adequate preventative maintenance was not performed on this system, resulting in failure, based on an inappropriate run to failure classification of this equipment.
The failure of this system resulted in a significant unplanned reduction in reactor power to 20 percent. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2007-04447.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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, Significance Determination Process, Phase 1 Worksheet, the finding has very low safety significance (Green) since the finding did not contribute to both the likelihood of a reactor trip and that mitigating equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007005 (pdf)
 
Mitigating Systems Significance:      Mar 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Improper Design Control for Evaluating Emergency Diesel Generator Turbocharger Combustion Air Pipe Stresses The inspectors identified a noncited violation of Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control,"
for failure to incorporate accurate design information into a calculation to determine emergency diesel generator turbocharger discharge combustion air pipe stresses. This resulted in pipe failure. Specifically, a calculation assumed nonconservative pipe wall thicknesses and process air temperatures, treated pipe end points as rigid anchors and failed to use stress intensification factors. This resulted in low calculated pipe stresses. With appropriately calculated pipe stress values, Entergy personnel could reasonably have been expected to adequately modify the combustion air piping to preclude subsequent failures. This issue was entered into the licensees corrective action program as Condition Report RBS-2008-2869.
This issue was determined to be more than minor because it affected the mitigating systems cornerstone objective and was similar to Manual Chapter 0612, Appendix E, Example 3.j because the errors were considered more than a minor calculation error in that the deficiency failed to identify the high pipe wall stresses that significantly reduced the overall allowable material strength margin. Later pipe and weld flaws developed at the intercooler adapter and turbocharger end connections that rendered the emergency diesel generator Division 2 inoperable. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that the issue was of very low safety significance (Green) because it did not screen as risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008002 (pdf)
Significance:      Sep 29, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Removing Control Building Chilled Water System from Service An NRC identified noncited violation of 10 CFR 50.65 (a)(4) was identified for the failure to assess and manage the increase in risk that may result from proposed maintenance activities on the control building chilled water system.
This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2007-03059.
Using NRC Manual Chapter 0612, Appendix B, Section 3, Item 5(h), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions that changed the outcome of the assessment. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time periods is less than 1.0E-6 Inspection Report# : 2007004 (pdf)
Barrier Integrity Significance:      Mar 29, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Reactor Startup Procedure Results in Six Control Rod Withdrawal Errors A self-revealing noncited violation of Technical Specification 5.4.1.a occurred when River Bend Station reactor operators failed to comply with General Operating Procedure GOP 000-1, Plant Start Up. Specifically operators withdrew six control rods two notches past the target out notch position specified in Reactivity Control Plan RCP 03. No fuel damage resulted from these errors. This issue was entered into the licensees corrective action program as Condition Report RBS-2008-2174.
 
This finding was more than minor because the finding affected the barrier integrity cornerstone attributes of configuration control and human performance and adversely impacts the cornerstones objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radio nuclide releases caused by accidents or events. The inspectors completed a Phase 1 significance determination using Manual Chapter 0609 Appendix A, Significance Determination Process Phase 1 screening worksheet, and determined the finding to be of very low safety significance (Green) because the performance issue only degraded the fuel cladding barrier. This finding had crosscutting aspects associated with human performance in the area of work practices in that the reactor operators failed to use self-check and peer-check during control rod reactivity manipulations (H.4.a).
Inspection Report# : 2008002 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Evaluate the Magnitude and Extent of Radiological Hazards Results in Personnel Contaminations A self-revealing noncited violation of 10 CFR 20.1501(a) was identified for failure to evaluate the magnitude and extent of radiological hazards associated with performing inspections of equipment in the containment building after a reactor trip on May 4, 2007. This failure resulted in six personnel contaminations and uptakes. Followup surveys identified contamination levels of 60 mRad/smear beta/gamma and up to 1300 dpm alpha. Air sample results determined a derived air concentration value of 44 for noble gas. The licensee has placed this event in the radiation protection continuing training program and entered it into their corrective action program as Condition Report CR-RBS-2007-1822.
This 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 in that the failure to evaluate the magnitude and extent of radiological hazards could cause unintentional dose to radiation workers. This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work control because the licensee did not communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2007005 (pdf)
Significance:      Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit and Radiation Worker Expectations A self-revealing noncited violation of Technical Specification 5.4.1 was identified for failure to follow radiation work permit instructions resulting in a worker entering a posted high radiation area without authorization. On April 20, 2007, an individual received an electronic alarming dosimeter dose rate alarm after entering a posted high radiation area. The individual was signed on to a radiation work permit that did not allow entry into a high radiation area. This violation was entered into licensees corrective action program as Condition Report CR-RBS-2007-1584.
This finding was greater than minor because it was associated with the occupational radiation safety cornerstone attribute of human performance and affected the cornerstone objective in that the failure to follow radiation work permit requirements could cause unintentional dose. This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did
 
not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work practices because the individual involved did not use proper self-checking and entered an area he was not authorized to enter.
Inspection Report# : 2007005 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post a Radiation Area An NRC-identified noncited violation of 10 CFR 20.1902(a) was identified for failure to conspicuously post a radiation area. Specifically, the inspector identified an entrance to a radiation area on the 90-foot elevation of the radwaste building that was accessible by a permanently installed ladder from the 65-foot elevation, which was not conspicuously posted as a radiation area. General area dose rates in the area were as high as 7 mrem/hour. This violation was entered into the licensees corrective action program as Condition Report CR-RBS-2007-4954.
This 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 in that the failure to post radiation areas could cause unintentional dose to radiation workers. This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work practices because radiation protection personnel did not adhere to management expectations regarding procedural compliance and following station procedures.
Inspection Report# : 2007005 (pdf)
Significance:      Sep 29, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Survey Following Containment Atmosphere Radiation Monitor Particulate Channel Alarms An NRC-identified noncited violation of 10 CFR 20.1501(a) was identified involving multiple failures to perform radiological surveys to evaluate radiological hazards following control room alarms of the Containment Atmosphere Radiation monitor particulate channel. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2007-04415.
This finding is more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of program and process, and affects the cornerstone objective to ensure the adequate protection of a workers health and safety from exposure to radiation because it could have resulted in workers being exposed to higher radiation levels. When processed through the Occupational Radiation Safety Significance Determination Process, the finding is determined to be of very low safety significance because it is not an as low as is reasonably achievable finding, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised.
The finding has a crosscutting aspect in the area of human performance, specifically the work control component, because the licensee failed to appropriately coordinate work activities by incorporating actions to address the impact of the work on different job activities and the need for work groups to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance (H.3(b)).
Inspection Report# : 2007004 (pdf)
Significance:      Jul 13, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Conspicuously Post Radiation Areas The team identified a noncited violation of 10 CFR 20.1902(a) because the licensee failed to post radiation areas in the radwaste building with a conspicuous sign or signs bearing the radiation symbol and the words Caution, Radiation Area. The licensee posted radiation area signs only at the entrances to the different elevations of the building, instead of at the discrete radiation areas, even though most of the radwaste building was not a radiation area.
 
Dose rates in unposted radiation areas were as high as 15 millirems per hour. As corrective action, the licensee posted the discrete areas. Additional corrective action is still being evaluated.
The finding was greater than minor because it was associated with one of the cornerstone attributes (exposure control and monitoring) and the finding affected the Occupational Radiation Safety cornerstone objective, in that, uninformed workers could unknowingly accrue additional radiation dose. Using the Occupational Radiation Safety Significance Determination Process, the team determined that the finding was of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Also, this finding had a cross-cutting aspect in the area of human performance and component of work control because the licensee did not coordinate work activities by incorporating actions to address the need to keep personnel apprised of plant conditions that may affect work activities.
Inspection Report# : 2007010 (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
 
River Bend 1 3Q/2008 Plant Inspection Findings Initiating Events Significance:        Jun 28, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment While Troubleshooting Results in Unanticipated Reactor Power Oscillations The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) when operators failed to perform an adequate risk assessment associated with a reactor start-up while performing troubleshooting, and during maintenance activities on the main turbine electro hydraulic control system. This resulted in unanticipated oscillations in reactor power and pressure. The licensee entered this issue into their corrective action program as Condition Report RBS-2008-4284.
Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), this finding is more than minor because Entergys risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. The risk assessment also failed to consider emergent maintenance activities that could increase the likelihood of initiating events. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component of decision making because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
Inspection Report# : 2008003 (pdf)
Significance:        Mar 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Internal Operating Experience Not Used to Prevent Recurrence of Reactor Recirculation FCV Runbacks The inspectors identified a noncited violation of Technical Specification 5.4.1.a for an inadequate procedure for securing a reactor feedwater pump. Specifically, the licensee failed to incorporate internal operating experience into the procedure. As a result, a reactor recirculation flow control valve runback resulting from a known reactor vessel water level loop tolerance issue recurred, resulting in an unplanned power reduction. This issue was entered into the licensees corrective action program as Condition Report RBS-2007-4749.
The finding is more than minor since it affects the human performance area 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. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding has very low safety significance since it did not contribute to both the likelihood of a reactor scram and the likelihood that mitigating equipment would not have been available.
Inspection Report# : 2008002 (pdf)
Significance:        Mar 29, 2008 Identified By: Self-Revealing Item Type: FIN Finding Condensate Demineralizer Tank Liner Failure A self-revealing finding was identified for the failure to properly repair condensate Demineralizer 1E tank liner prior to returning it to service.
As a result, failure of the liner resulted in approximately 20,000 gallons of radiological contaminated condensate being spilled from the manway flange. Operations lowered reactor power from 90 percent to 82 percent to conserve condensate system inventory. This issue was entered into the licensees corrective action program as Condition Report RBS-2007-5440.
The finding is greater than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was considered to be a transient initiator contributor which contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available and, therefore, screened to Phase 2. Using the Phase 2 worksheets, the inspectors assumed that successful recovery of the condensate system from the leak was highly likely and determined the finding to be of very low safety significance. This finding has crosscutting aspects associated with human performance in the area of resources in that a complete, accurate, and up-to-date work package was not available to assure nuclear safety [H.2(c)].
Inspection Report# : 2008002 (pdf)
 
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation RPS Terminal Board Loose Connection Results in a Reactor Scram A self-revealing noncited violation of 10 CFR Part 50 Appendix B, Criterion V was identified involving the failure to adequately torque reactor protection system electrical terminal board connections during initial construction. This failure resulted in a loose terminal connection causing thermal degradation that subsequently resulted in an automatic reactor scram during average power range monitor surveillance testing. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2007-04264.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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, Significance Determination Process, Phase 1 Worksheet, the finding was determined to have very low safety significance (Green) because the finding did not contribute to both the likelihood of a reactor trip and that mitigating equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Preventive Maintenance Strategy Results in a Breaker's Failure to Promptly Open Due to Hardened Grease Results in a Complicated Reactor Scram A self-revealing Green noncited violation of 10 CFR 50.65(A)(3) was identified for failure to incorporate internal and external operating experience into preventive maintenance activities to prevent industry known electrical circuit breaker deficiencies. Specifically, inadequate breaker maintenance, leading to grease hardening degradation, resulted in inadequate electrical fault protection on November 7, 2007. The failure to adequately isolate the electrical fault resulted in a complicated reactor scram involving the loss of the main condenser and reactor feedwater. The licensee entered this into their corrective action program as CR-RBS-2007-04922.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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. The inspectors evaluated the finding using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations." The finding required a Phase 2 analysis because the finding contributed to the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. A senior reactor analyst estimated the risk of the subject finding using the Risk-Informed Inspection Notebook for River Bend Station, Unit 1, Revision 2.1a. The analyst determined the finding was of very low safety significance.
This finding has a crosscutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Preventative Maintenance Results in a Plant Down Power A self-revealing finding was identified for failure to perform adequate preventive maintenance for control panels associated with providing make up water to the circulating water system. Adequate preventative maintenance was not performed on this system, resulting in failure, based on an inappropriate run to failure classification of this equipment. The failure of this system resulted in a significant unplanned reduction in reactor power to 20 percent. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2007-04447.
The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance 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, Significance Determination Process, Phase 1 Worksheet, the finding has very low safety significance (Green) since the finding did not contribute to both the likelihood of a reactor trip and that mitigating equipment or functions would not be available following a reactor trip.
Inspection Report# : 2007005 (pdf)
Mitigating Systems
 
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Eight Examples of a Failure to Meet 10 CFR Part 50, Appendix B, "Design Control" The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with eight examples.
* Example 1: Non-conservative inputs and assumptions used without adequate technical justification to evaluate the minimum terminal voltage and actuator output torque for safety-related motor operated valves. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03339.
* Example 2: Failure to perform a conservative analysis to ensure that Technical Specification Setpoints were adequate. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03911.
* Example 3: Non-conservative inputs and methodologies used in calculating control circuit voltages to safety-related 480V motor operated valves motor-operated valve and motors that would be required to operate for mitigation of design bases events. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03858.
* Example 4: Failure to evaluate E12-MOV-F042A, residual heat removal injection valve, and E12-MOV-F064A, residual heat removal minimum flow valve, to verify adequate voltage would be available to operate the associated 120VAC control circuit devices. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03641
* Example 5: Inadequate design basis documentation for hydrogen concentration control in the Division I and II Battery Rooms in the control building. After identification, the licensee entered the issue into the corrective action program as Condition Reports CR-RBS-2008-02566 and CR-RBS-2008-03403.
* Example 6: Failure to ensure design basis information for safety related 125VDC batteries was controlled and correctly translated into procedures and instructions. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03659.
* Example 7: Failure to maintain adequate design basis calculations for ultimate heat sink loading. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-3712.
* Example 8: Failure to account for the technical specification allowed emergency diesel generator frequency variation in the diesel loading calculation. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03556.
The examples associated with this finding were more than minor per Manual Chapter 612, Appendix E, Appendix E, Examples of Minor Issues, Example 3j, in that each example resulted in a condition where there was reasonable doubt on the operability of a system or component. The finding 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined each example was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality.
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Recalculate Suppression Pool Peak Temperature Rseponse The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that, design control measures for verifying the adequacy of design were not implemented. Specifically, the licensee did not recalculate suppression pool peak temperature response when a more severe single failure condition was identified. In response, the licensee entered this issue in the corrective action program as Condition Report CR-RBS-2008-03661 and determined that suppression pool peak temperature response was acceptable.
The 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality of the suppression pool. The finding had a cross-cutting aspect in the area of
 
problem identification and resolution because the licensee initiated a corrective action program action to re-evaluate long-term suppression pool peak temperature performance but closed the action without its completion [P.1 (d)].
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Testing Programs for 4-kV Circuit Breakers, Class 1E Molded Case Circuit Breakers, and the Emergency Diesel Generators The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, with three examples. Specifically, the team identified that the licensee failed to develop and implement adequate testing programs for 4-kV circuit breakers, Class 1E molded-case circuit breakers, and the emergency diesel generators that met design or vendor requirements and recommendations. In response, the licensee entered these examples in the corrective action program as Condition Reports CR RBS-2008-04379, CR-RBS-2008-3634, CR-RBS-2008-3676 and CR-RBS-2008-3701 and determined there was no loss of safety function for the affected components.
The examples associated with this finding were more than minor because they were associated with the equipment 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined each example was of very low safety significance (Green) because it did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather.
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Implementation of Temporary Installation Procedure The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for five examples of failure to follow the requirements of ADM-0073 Temporary Installation Guidelines" during the installation of modifications to the plant. Specifically, four modifications were installed in the plant that did not meet the criteria of a temporary installation and one was not removed when no longer needed, as required by the procedure. After identification, the licensee entered the issue into the corrective action program as CR-RBS-2008-3410.
Although the team considered each of the above examples minor in significance, the team determined that this finding, which was associated with design control attribute of the Mitigating Systems cornerstone, was more than minor per Manual Chapter 612, Appendix E, Examples of Minor Issues, Example 4a. The finding involved multiple examples of failure to follow licensee procedural requirements and if left uncorrected it could result in design modifications to the plant that were not properly evaluated, controlled, documented and installed.
Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect associated with resources in the human performance area because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, those necessary for maintaining long term plant safety by maintenance of design margins, minimization of long-standing equipment issues, minimizing preventative maintenance deferrals, and ensuring maintenance and engineering backlogs which were low enough to support safety [H.2 (a)].
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Implementation of Operability Determination Procedure The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the failure to follow procedures to evaluate conditions adverse to quality for impacts on the
 
operability of safety-related equipment. Specifically, the licensee did not assess the impact on operability of previous steam leaks and motor-stall events on the corrosion of magnesium-rotors in safety-related motor-operated valves. The licensee entered this issue into the corrective action program as Condition Reports CR-RBS-2008-3713 and CR-RBS-2008-3766.
The finding was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of safety-related motor-operated valves to respond to initiating events to prevent undesirable consequences. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The cause of the finding had crosscutting aspects associated with the corrective action program in the problem identification and resolution area because the licensee did not thoroughly evaluate the problems with magnesium-rotor corrosion including the extent of the condition and operability impact [P.1(c)].
Inspection Report# : 2008006 (pdf)
Significance:        Jun 28, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Removing Control Building Chilled Water System from Service The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) when operators failed to perform an adequate risk assessment while the Division 1 control building chilled water and control building air conditioning systems were unavailable. Specifically, the inspectors identified that licensee personnel nonconservatively evaluated the on-line risk as Green instead of Yellow. This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis evaluation. The licensee entered this issue into their corrective action program as Condition Report RBS-2008-2687.
Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected period is less than 1.0E-
: 6. This finding has a crosscutting aspect in the area of problem identification and resolution component of operating experience because Entergy did not systematically communicate to affected internal stakeholders in a timely manner relevant internal operating experience [P.2 (a)].
Inspection Report# : 2008003 (pdf)
Significance:        Mar 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Improper Design Control for Evaluating Emergency Diesel Generator Turbocharger Combustion Air Pipe Stresses The inspectors identified a noncited violation of Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to incorporate accurate design information into a calculation to determine emergency diesel generator turbocharger discharge combustion air pipe stresses.
This resulted in pipe failure. Specifically, a calculation assumed nonconservative pipe wall thicknesses and process air temperatures, treated pipe end points as rigid anchors and failed to use stress intensification factors. This resulted in low calculated pipe stresses. With appropriately calculated pipe stress values, Entergy personnel could reasonably have been expected to adequately modify the combustion air piping to preclude subsequent failures. This issue was entered into the licensees corrective action program as Condition Report RBS-2008-2869.
This issue was determined to be more than minor because it affected the mitigating systems cornerstone objective and was similar to Manual Chapter 0612, Appendix E, Example 3.j because the errors were considered more than a minor calculation error in that the deficiency failed to identify the high pipe wall stresses that significantly reduced the overall allowable material strength margin. Later pipe and weld flaws developed at the intercooler adapter and turbocharger end connections that rendered the emergency diesel generator Division 2 inoperable.
Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that the issue was of very low safety significance (Green) because it did not screen as risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008002 (pdf)
Barrier Integrity Significance:        Aug 26, 2008 Identified By: NRC
 
Item Type: NCV NonCited Violation Inadequate/Untimely Corrective Action for Failure of Magnesium-Rotor Motor-Operated Valves The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, "Corrective Action," for failure to promptly identify magnesium-rotor motor-operated valve degradation. Specifically, the licensee did not identify magnesium-rotor degradation in May 2007 after failure of Valve B21-MOV-FO65A, Reactor Inlet Heater A Outboard Motor Operated Isolation Valve, until after failure of Valve B21-MOV-FO98C, Main Steam Shutoff Valve, in September 2007. The licensee entered this issue into the corrective action program as Condition Reports CR-RBS-2008-3713 and CR-RBS-2008-3766.
This finding was more than minor because Valve B21-MOV-FO98C was associated with the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that the physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. Inspection Manual chapter 0609 Appendix H, "Containment Integrity Significance Determination Process," Table 4.1, indicated that the Main Steam Shutoff Valves do not impact large early release frequency. Based on the results of the Appendix H analysis, the finding was determined to have very low safety significance. This finding had cross-cutting aspects associated with decision-making in the human performance area in that the licensee did not use conservative assumptions in decision-making regarding the likelihood of magnesium-rotor degradation in motor-operated valves [H.1 (b)].
Inspection Report# : 2008006 (pdf)
Significance:        Mar 29, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Reactor Startup Procedure Results in Six Control Rod Withdrawal Errors A self-revealing noncited violation of Technical Specification 5.4.1.a occurred when River Bend Station reactor operators failed to comply with General Operating Procedure GOP 000-1, Plant Start Up. Specifically operators withdrew six control rods two notches past the target out notch position specified in Reactivity Control Plan RCP-15-03. No fuel damage resulted from these errors. This issue was entered into the licensees corrective action program as Condition Report RBS-2008-2174.
This finding was more than minor because the finding affected the barrier integrity cornerstone attributes of configuration control and human performance and adversely impacts the cornerstones objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radio nuclide releases caused by accidents or events. The inspectors completed a Phase 1 significance determination using Manual Chapter 0609 Appendix A, Significance Determination Process Phase 1 screening worksheet, and determined the finding to be of very low safety significance (Green) because the performance issue only degraded the fuel cladding barrier. This finding had crosscutting aspects associated with human performance in the area of work practices in that the reactor operators failed to use self-check and peer-check during control rod reactivity manipulations (H.4.a).
Inspection Report# : 2008002 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Jun 28, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow a Radiation Protection Procedure The inspector reviewed a self-revealing noncited violation of Technical Specification 5.4.1 which resulted from workers failing to follow a radiation protection procedure. On January 13, 2008, three workers attempted to exit the radiologically controlled area and alarmed the personnel contamination monitors. The workers were removing tubes from the Water Box B. The licensee determined radiation protection staff did not follow the radiation work permit planning procedure to use representative radiological surveys for the work performed. The radiation work permit planning did not include previous water box internal and other related surveys which would correspond to the removal of the water box tubes. The licensees investigation found that the contamination levels on the tubes were as high as 150,000 disintegrations per minute per 100 cm2. The licensee revised the radiation work permit to include the actual working conditions and appropriate personnel protective equipment.
Workers failing to follow a radiation protection procedure is a performance deficiency. The finding is greater than minor because, if left uncorrected, the deficiency would become a more radiologically significant safety concern resulting in additional workers unplanned,
 
unintended dose as work continued to be performed under an inadequate radiation work permit. Since this issue involved workers unplanned and unintended dose, the Occupational Radiation Safety Significance Determination Process was used to determine the safety significance.
The inspector determined the finding had very low safety significance because: (1) it was not an ALARA finding, (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 control, because the radiation protection staff did not plan work activities consistent with radiological safety by incorporating risk insights and job site conditions of the actual work to be performed during the radiation work permit planning [H.3(a)].
Inspection Report# : 2008003 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Evaluate the Magnitude and Extent of Radiological Hazards Results in Personnel Contaminations A self-revealing noncited violation of 10 CFR 20.1501(a) was identified for failure to evaluate the magnitude and extent of radiological hazards associated with performing inspections of equipment in the containment building after a reactor trip on May 4, 2007. This failure resulted in six personnel contaminations and uptakes. Followup surveys identified contamination levels of 60 mRad/smear beta/gamma and up to 1300 dpm alpha. Air sample results determined a derived air concentration value of 44 for noble gas. The licensee has placed this event in the radiation protection continuing training program and entered it into their corrective action program as Condition Report CR-RBS-2007-1822.
This 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 in that the failure to evaluate the magnitude and extent of radiological hazards could cause unintentional dose to radiation workers. This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work control because the licensee did not communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit and Radiation Worker Expectations A self-revealing noncited violation of Technical Specification 5.4.1 was identified for failure to follow radiation work permit instructions resulting in a worker entering a posted high radiation area without authorization. On April 20, 2007, an individual received an electronic alarming dosimeter dose rate alarm after entering a posted high radiation area. The individual was signed on to a radiation work permit that did not allow entry into a high radiation area. This violation was entered into licensees corrective action program as Condition Report CR-RBS-2007-1584.
This finding was greater than minor because it was associated with the occupational radiation safety cornerstone attribute of human performance and affected the cornerstone objective in that the failure to follow radiation work permit requirements could cause unintentional dose. This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work practices because the individual involved did not use proper self-checking and entered an area he was not authorized to enter.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post a Radiation Area An NRC-identified noncited violation of 10 CFR 20.1902(a) was identified for failure to conspicuously post a radiation area. Specifically, the inspector identified an entrance to a radiation area on the 90-foot elevation of the radwaste building that was accessible by a permanently installed ladder from the 65-foot elevation, which was not conspicuously posted as a radiation area. General area dose rates in the area were as high as 7 mrem/hour. This violation was entered into the licensees corrective action program as Condition Report CR-RBS-2007-4954.
This 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 in that the failure to post radiation areas could cause unintentional dose to radiation workers.
This finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding had a crosscutting aspect in the area of human performance related to the component of work practices because radiation protection personnel did not adhere to management expectations regarding procedural
 
compliance and following station procedures.
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
 
River Bend 1 4Q/2008 Plant Inspection Findings Initiating Events Significance:        Sep 27, 2008 Identified By: Self-Revealing Item Type: FIN Finding Turbine Building Siding Failure Below Design Specifications A self-revealing finding was identified for wind induced turbine building siding failure that occurred significantly below design specified stress levels as a result of design and installation deficiencies. This resulted in a forced outage to repair transformer damage and to repair the turbine building siding. The licensee missed prior opportunities to identify turbine building siding design and installation deficiencies following damaging wind events in 1992 and 2005. The licensee entered this issue into the corrective action program as Condition Report CR-RBS-2008-5176.
This finding is more than minor because it is associated with the protection against external factors attribute (wind and grid stability) of the initiating events cornerstone and affected 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.
The inspectors evaluated the significance of this finding using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined it to be of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.
Inspection Report# : 2008004 (pdf)
Significance:        Jun 28, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment While Troubleshooting Results in Unanticipated Reactor Power Oscillations The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) when operators failed to perform an adequate risk assessment associated with a reactor start-up while performing troubleshooting, and during maintenance activities on the main turbine electro hydraulic control system. This resulted in unanticipated oscillations in reactor power and pressure. The licensee entered this issue into their corrective action program as Condition Report RBS-2008-4284.
Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), this finding is more than minor because Entergys risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. The risk assessment also failed to consider emergent maintenance activities that could increase the likelihood of initiating events. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component of decision making because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
Inspection Report# : 2008003 (pdf)
Significance:        Mar 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Internal Operating Experience Not Used to Prevent Recurrence of Reactor Recirculation FCV Runbacks
 
The inspectors identified a noncited violation of Technical Specification 5.4.1.a for an inadequate procedure for securing a reactor feedwater pump. Specifically, the licensee failed to incorporate internal operating experience into the procedure. As a result, a reactor recirculation flow control valve runback resulting from a known reactor vessel water level loop tolerance issue recurred, resulting in an unplanned power reduction. This issue was entered into the licensees corrective action program as Condition Report RBS-2007-4749.
The finding is more than minor since it affects the human performance area 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. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding has very low safety significance since it did not contribute to both the likelihood of a reactor scram and the likelihood that mitigating equipment would not have been available.
Inspection Report# : 2008002 (pdf)
Significance:        Mar 29, 2008 Identified By: Self-Revealing Item Type: FIN Finding Condensate Demineralizer Tank Liner Failure A self-revealing finding was identified for the failure to properly repair condensate Demineralizer 1E tank liner prior to returning it to service. As a result, failure of the liner resulted in approximately 20,000 gallons of radiological contaminated condensate being spilled from the manway flange. Operations lowered reactor power from 90 percent to 82 percent to conserve condensate system inventory. This issue was entered into the licensees corrective action program as Condition Report RBS-2007-5440.
The finding is greater than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was considered to be a transient initiator contributor which contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available and, therefore, screened to Phase 2. Using the Phase 2 worksheets, the inspectors assumed that successful recovery of the condensate system from the leak was highly likely and determined the finding to be of very low safety significance. This finding has crosscutting aspects associated with human performance in the area of resources in that a complete, accurate, and up-to-date work package was not available to assure nuclear safety
[H.2(c)].
Inspection Report# : 2008002 (pdf)
Mitigating Systems Significance:        Sep 27, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Staging the Station Blackout Diesel Generator during Severe Weather The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving the failure to have an adequate procedure to ensure the availability of on site emergency ac power sources following the four-hour coping period of a postulated station blackout. Specifically, station procedures did not ensure that the station blackout diesel generator would be reliably deployed to fulfill its intended function during sustained high winds. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-5050.
This finding is more than minor because it is associated with the protection against external factors attribute (wind and grid stability) 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
 
inspectors evaluated the significance of this finding using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined it to be of very low safety significance because it did not result in an actual loss of safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 27, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Corrective Actions Results in Multiple Failures of Standby Service Water Switchgear Room Ventilation Fans A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to take adequate corrective actions in response to a condition adverse to quality resulting in repetitive failures of the standby service water switchgear room ventilation fans. Following failure of the switchgear fans in July 2008, the licensee found that inappropriate flow switch settings on the fans had been identified in a condition report in October 1999, but no actions had been taken to correct the condition. Subsequently, more failures of the standby service water switchgear room ventilation fans occurred, including nineteen in the past three and one half years, many of which were attributed to flow switch issues. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-5761.
The finding was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone, and it directly affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to preclude 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 the condition did not result in an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. This finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee failed to maintain long term plant safety by minimization of long standing equipment issues [H.2(a)].
Inspection Report# : 2008004 (pdf)
Significance:        Sep 27, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Transformer Yard Maintenance While Shut Down The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the licensees failure to assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, while conducting maintenance in the transformer yard during severe weather with high pressure core spray inoperable, the licensee did not assess the affects on the shutdown risk. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-05383.
The inspectors determined this finding was more than minor since it was similar to Manual Chapter 0612, Appendix E, Example 7.e, and since it caused the licensees risk model to change from a Green to Yellow risk window. In accordance with NRC Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management, the inspectors requested that a senior reactor analyst evaluate the risk of this condition. The analyst determined that this finding was of very low risk significance because the associated risk deficit was less than 1.0E-6.
Inspection Report# : 2008004 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation
 
Eight Examples of a Failure to Meet 10 CFR Part 50, Appendix B, "Design Control" The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with eight examples.
* Example 1: Non-conservative inputs and assumptions used without adequate technical justification to evaluate the minimum terminal voltage and actuator output torque for safety-related motor operated valves. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03339.
* Example 2: Failure to perform a conservative analysis to ensure that Technical Specification Setpoints were adequate. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03911.
* Example 3: Non-conservative inputs and methodologies used in calculating control circuit voltages to safety-related 480V motor operated valves motor-operated valve and motors that would be required to operate for mitigation of design bases events. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03858.
* Example 4: Failure to evaluate E12-MOV-F042A, residual heat removal injection valve, and E12-MOV-F064A, residual heat removal minimum flow valve, to verify adequate voltage would be available to operate the associated 120VAC control circuit devices. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03641
* Example 5: Inadequate design basis documentation for hydrogen concentration control in the Division I and II Battery Rooms in the control building. After identification, the licensee entered the issue into the corrective action program as Condition Reports CR-RBS-2008-02566 and CR-RBS-2008-03403.
* Example 6: Failure to ensure design basis information for safety related 125VDC batteries was controlled and correctly translated into procedures and instructions. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03659.
* Example 7: Failure to maintain adequate design basis calculations for ultimate heat sink loading. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-3712.
* Example 8: Failure to account for the technical specification allowed emergency diesel generator frequency variation in the diesel loading calculation. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03556.
The examples associated with this finding were more than minor per Manual Chapter 612, Appendix E, Appendix E, Examples of Minor Issues, Example 3j, in that each example resulted in a condition where there was reasonable doubt on the operability of a system or component. The finding 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined each example was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality.
Inspection Report# : 2008006 (pdf)
Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Recalculate Suppression Pool Peak Temperature Rseponse The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50,
 
Appendix B, Criterion III, Design Control, in that, design control measures for verifying the adequacy of design were not implemented. Specifically, the licensee did not recalculate suppression pool peak temperature response when a more severe single failure condition was identified. In response, the licensee entered this issue in the corrective action program as Condition Report CR-RBS-2008-03661 and determined that suppression pool peak temperature response was acceptable.
The 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality of the suppression pool. The finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee initiated a corrective action program action to re-evaluate long-term suppression pool peak temperature performance but closed the action without its completion [P.1 (d)].
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Testing Programs for 4-kV Circuit Breakers, Class 1E Molded Case Circuit Breakers, and the Emergency Diesel Generators The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, with three examples. Specifically, the team identified that the licensee failed to develop and implement adequate testing programs for 4-kV circuit breakers, Class 1E molded-case circuit breakers, and the emergency diesel generators that met design or vendor requirements and recommendations. In response, the licensee entered these examples in the corrective action program as Condition Reports CR RBS-2008-04379, CR-RBS-2008-3634, CR-RBS-2008-3676 and CR-RBS-2008-3701 and determined there was no loss of safety function for the affected components.
The examples associated with this finding were more than minor because they were associated with the equipment 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined each example was of very low safety significance (Green) because it did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather.
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Implementation of Temporary Installation Procedure The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for five examples of failure to follow the requirements of ADM-0073 Temporary Installation Guidelines" during the installation of modifications to the plant.
 
Specifically, four modifications were installed in the plant that did not meet the criteria of a temporary installation and one was not removed when no longer needed, as required by the procedure. After identification, the licensee entered the issue into the corrective action program as CR-RBS-2008-3410.
Although the team considered each of the above examples minor in significance, the team determined that this finding, which was associated with design control attribute of the Mitigating Systems cornerstone, was more than minor per Manual Chapter 612, Appendix E, Examples of Minor Issues, Example 4a. The finding involved multiple examples of failure to follow licensee procedural requirements and if left uncorrected it could result in design modifications to the plant that were not properly evaluated, controlled, documented and installed. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect associated with resources in the human performance area because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, those necessary for maintaining long term plant safety by maintenance of design margins, minimization of long-standing equipment issues, minimizing preventative maintenance deferrals, and ensuring maintenance and engineering backlogs which were low enough to support safety [H.2 (a)].
Inspection Report# : 2008006 (pdf)
Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Implementation of Operability Determination Procedure The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the failure to follow procedures to evaluate conditions adverse to quality for impacts on the operability of safety-related equipment. Specifically, the licensee did not assess the impact on operability of previous steam leaks and motor-stall events on the corrosion of magnesium-rotors in safety-related motor-operated valves. The licensee entered this issue into the corrective action program as Condition Reports CR-RBS-2008-3713 and CR-RBS-2008-3766.
The finding was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of safety-related motor-operated valves to respond to initiating events to prevent undesirable consequences. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The cause of the finding had crosscutting aspects associated with the corrective action program in the problem identification and resolution area because the licensee did not thoroughly evaluate the problems with magnesium-rotor corrosion including the extent of the condition and operability impact [P.1(c)].
Inspection Report# : 2008006 (pdf)
Significance:      Jun 28, 2008 Identified By: NRC
 
Item Type: NCV NonCited Violation Inadequate Risk Assessment for Removing Control Building Chilled Water System from Service The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) when operators failed to perform an adequate risk assessment while the Division 1 control building chilled water and control building air conditioning systems were unavailable. Specifically, the inspectors identified that licensee personnel nonconservatively evaluated the on-line risk as Green instead of Yellow. This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis evaluation. The licensee entered this issue into their corrective action program as Condition Report RBS-2008-2687.
Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected period is less than 1.0E-6. This finding has a crosscutting aspect in the area of problem identification and resolution component of operating experience because Entergy did not systematically communicate to affected internal stakeholders in a timely manner relevant internal operating experience [P.2(a)].
Inspection Report# : 2008003 (pdf)
Significance:      Mar 29, 2008 Identified By: NRC Item Type: NCV NonCited Violation Improper Design Control for Evaluating Emergency Diesel Generator Turbocharger Combustion Air Pipe Stresses The inspectors identified a noncited violation of Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control,"
for failure to incorporate accurate design information into a calculation to determine emergency diesel generator turbocharger discharge combustion air pipe stresses. This resulted in pipe failure. Specifically, a calculation assumed nonconservative pipe wall thicknesses and process air temperatures, treated pipe end points as rigid anchors and failed to use stress intensification factors. This resulted in low calculated pipe stresses. With appropriately calculated pipe stress values, Entergy personnel could reasonably have been expected to adequately modify the combustion air piping to preclude subsequent failures. This issue was entered into the licensees corrective action program as Condition Report RBS-2008-2869.
This issue was determined to be more than minor because it affected the mitigating systems cornerstone objective and was similar to Manual Chapter 0612, Appendix E, Example 3.j because the errors were considered more than a minor calculation error in that the deficiency failed to identify the high pipe wall stresses that significantly reduced the overall allowable material strength margin. Later pipe and weld flaws developed at the intercooler adapter and turbocharger end connections that rendered the emergency diesel generator Division 2 inoperable. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet, the inspectors determined that the issue was of very low safety significance (Green) because it did not screen as risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008002 (pdf)
Barrier Integrity Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate/Untimely Corrective Action for Failure of Magnesium-Rotor Motor-Operated Valves The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, "Corrective Action," for failure to promptly identify magnesium-rotor motor-operated valve degradation. Specifically, the licensee did not identify magnesium-rotor degradation in May 2007 after failure
 
of Valve B21-MOV-FO65A, Reactor Inlet Heater A Outboard Motor Operated Isolation Valve, until after failure of Valve B21-MOV-FO98C, Main Steam Shutoff Valve, in September 2007. The licensee entered this issue into the corrective action program as Condition Reports CR-RBS-2008-3713 and CR-RBS-2008-3766.
This finding was more than minor because Valve B21-MOV-FO98C was associated with the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that the physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. Inspection Manual chapter 0609 Appendix H, "Containment Integrity Significance Determination Process," Table 4.1, indicated that the Main Steam Shutoff Valves do not impact large early release frequency. Based on the results of the Appendix H analysis, the finding was determined to have very low safety significance. This finding had cross-cutting aspects associated with decision-making in the human performance area in that the licensee did not use conservative assumptions in decision-making regarding the likelihood of magnesium-rotor degradation in motor-operated valves [H.1 (b)].
Inspection Report# : 2008006 (pdf)
Significance:      Mar 29, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Reactor Startup Procedure Results in Six Control Rod Withdrawal Errors A self-revealing noncited violation of Technical Specification 5.4.1.a occurred when River Bend Station reactor operators failed to comply with General Operating Procedure GOP 000-1, Plant Start Up. Specifically operators withdrew six control rods two notches past the target out notch position specified in Reactivity Control Plan RCP 03. No fuel damage resulted from these errors. This issue was entered into the licensees corrective action program as Condition Report RBS-2008-2174.
This finding was more than minor because the finding affected the barrier integrity cornerstone attributes of configuration control and human performance and adversely impacts the cornerstones objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radio nuclide releases caused by accidents or events. The inspectors completed a Phase 1 significance determination using Manual Chapter 0609 Appendix A, Significance Determination Process Phase 1 screening worksheet, and determined the finding to be of very low safety significance (Green) because the performance issue only degraded the fuel cladding barrier. This finding had crosscutting aspects associated with human performance in the area of work practices in that the reactor operators failed to use self-check and peer-check during control rod reactivity manipulations (H.4.a).
Inspection Report# : 2008002 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Jun 28, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow a Radiation Protection Procedure The inspector reviewed a self-revealing noncited violation of Technical Specification 5.4.1 which resulted from workers failing to follow a radiation protection procedure. On January 13, 2008, three workers attempted to exit the
 
radiologically controlled area and alarmed the personnel contamination monitors. The workers were removing tubes from the Water Box B. The licensee determined radiation protection staff did not follow the radiation work permit planning procedure to use representative radiological surveys for the work performed. The radiation work permit planning did not include previous water box internal and other related surveys which would correspond to the removal of the water box tubes. The licensees investigation found that the contamination levels on the tubes were as high as 150,000 disintegrations per minute per 100 cm2. The licensee revised the radiation work permit to include the actual working conditions and appropriate personnel protective equipment.
Workers failing to follow a radiation protection procedure is a performance deficiency. The finding is greater than minor because, if left uncorrected, the deficiency would become a more radiologically significant safety concern resulting in additional workers unplanned, unintended dose as work continued to be performed under an inadequate radiation work permit. Since this issue involved workers unplanned and unintended dose, the Occupational Radiation Safety Significance Determination Process was used to determine the safety significance. The inspector determined the finding had very low safety significance because: (1) it was not an ALARA finding, (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 control, because the radiation protection staff did not plan work activities consistent with radiological safety by incorporating risk insights and job site conditions of the actual work to be performed during the radiation work permit planning [H.3 (a)].
Inspection Report# : 2008003 (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
 
River Bend 1 1Q/2009 Plant Inspection Findings Initiating Events Significance:        Sep 27, 2008 Identified By: Self-Revealing Item Type: FIN Finding Turbine Building Siding Failure Below Design Specifications A self-revealing finding was identified for wind induced turbine building siding failure that occurred significantly below design specified stress levels as a result of design and installation deficiencies. This resulted in a forced outage to repair transformer damage and to repair the turbine building siding. The licensee missed prior opportunities to identify turbine building siding design and installation deficiencies following damaging wind events in 1992 and 2005. The licensee entered this issue into the corrective action program as Condition Report CR-RBS-2008-5176.
This finding is more than minor because it is associated with the protection against external factors attribute (wind and grid stability) of the initiating events cornerstone and affected 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.
The inspectors evaluated the significance of this finding using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined it to be of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.
Inspection Report# : 2008004 (pdf)
Significance:        Jun 28, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment While Troubleshooting Results in Unanticipated Reactor Power Oscillations The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) when operators failed to perform an adequate risk assessment associated with a reactor start-up while performing troubleshooting, and during maintenance activities on the main turbine electro hydraulic control system. This resulted in unanticipated oscillations in reactor power and pressure. The licensee entered this issue into their corrective action program as Condition Report RBS-2008-4284.
Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), this finding is more than minor because Entergys risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. The risk assessment also failed to consider emergent maintenance activities that could increase the likelihood of initiating events. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component of decision making because the licensee did not use conservative assumptions in decision making and adopt a requirement to demonstrate the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action [H.1(b)].
Inspection Report# : 2008003 (pdf)
Mitigating Systems
 
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment While the Control Building Chilled Water System was Removed from Service The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) involving the failure of operators to perform an adequate risk assessment while the Division 1 control building chilled water was unavailable. Specifically, the inspectors identified that licensee personnel non-conservatively evaluated the on-line risk as Green instead of Yellow. This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis evaluation. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-0862.
Using Inspection Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions regarding the unavailability of mitigating systems that put the plant in a higher risk category. Using Inspection Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component for work practices because Entergy personnel did not effectively follow procedures [H.4(b)].
Inspection Report# : 2009002 (pdf)
Significance:        Sep 27, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Staging the Station Blackout Diesel Generator during Severe Weather The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving the failure to have an adequate procedure to ensure the availability of on site emergency ac power sources following the four-hour coping period of a postulated station blackout. Specifically, station procedures did not ensure that the station blackout diesel generator would be reliably deployed to fulfill its intended function during sustained high winds. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-5050.
This finding is more than minor because it is associated with the protection against external factors attribute (wind and grid stability) 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 inspectors evaluated the significance of this finding using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined it to be of very low safety significance because it did not result in an actual loss of safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 27, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Corrective Actions Results in Multiple Failures of Standby Service Water Switchgear Room Ventilation Fans A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to take adequate corrective actions in response to a condition adverse to quality resulting in repetitive failures of the standby service water switchgear room ventilation fans. Following failure of the switchgear fans in July 2008, the licensee found that inappropriate flow switch settings on the fans had been identified in a condition report in October 1999, but no actions had been taken to correct the condition. Subsequently, more failures of the standby service water switchgear room ventilation fans occurred, including nineteen in the past three and one half years, many of which were attributed to flow switch issues. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-5761.
 
The finding was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone, and it directly affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to preclude 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 the condition did not result in an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. This finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee failed to maintain long term plant safety by minimization of long standing equipment issues [H.2(a)].
Inspection Report# : 2008004 (pdf)
Significance:      Sep 27, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Transformer Yard Maintenance While Shut Down The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the licensees failure to assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, while conducting maintenance in the transformer yard during severe weather with high pressure core spray inoperable, the licensee did not assess the affects on the shutdown risk. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-05383.
The inspectors determined this finding was more than minor since it was similar to Manual Chapter 0612, Appendix E, Example 7.e, and since it caused the licensees risk model to change from a Green to Yellow risk window. In accordance with NRC Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management, the inspectors requested that a senior reactor analyst evaluate the risk of this condition. The analyst determined that this finding was of very low risk significance because the associated risk deficit was less than 1.0E-6.
Inspection Report# : 2008004 (pdf)
Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Eight Examples of a Failure to Meet 10 CFR Part 50, Appendix B, "Design Control" The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with eight examples.
* Example 1: Non-conservative inputs and assumptions used without adequate technical justification to evaluate the minimum terminal voltage and actuator output torque for safety-related motor operated valves. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03339.
* Example 2: Failure to perform a conservative analysis to ensure that Technical Specification Setpoints were adequate. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03911.
* Example 3: Non-conservative inputs and methodologies used in calculating control circuit voltages to safety-related 480V motor operated valves motor-operated valve and motors that would be required to operate for mitigation of design bases events. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03858.
* Example 4: Failure to evaluate E12-MOV-F042A, residual heat removal injection valve, and E12-MOV-F064A, residual heat removal minimum flow valve, to verify adequate voltage would be available to operate the associated 120VAC control circuit devices. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03641
* Example 5: Inadequate design basis documentation for hydrogen concentration control in the Division I and II Battery Rooms in the control building. After identification, the licensee entered the issue into the corrective action program as Condition Reports CR-RBS-2008-02566 and CR-RBS-2008-03403.
* Example 6: Failure to ensure design basis information for safety related 125VDC batteries was controlled and correctly translated into procedures and instructions. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03659.
* Example 7: Failure to maintain adequate design basis calculations for ultimate heat sink loading. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-3712.
* Example 8: Failure to account for the technical specification allowed emergency diesel generator frequency variation in the diesel loading calculation. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03556.
The examples associated with this finding were more than minor per Manual Chapter 612, Appendix E, Appendix E, Examples of Minor Issues, Example 3j, in that each example resulted in a condition where there was reasonable doubt on the operability of a system or component. The finding 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined each example was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality.
Inspection Report# : 2008006 (pdf)
Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Recalculate Suppression Pool Peak Temperature Rseponse The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that, design control measures for verifying the adequacy of design were not implemented. Specifically, the licensee did not recalculate suppression pool peak temperature response when a more severe single failure condition was identified. In response, the licensee entered this issue in the corrective action program as Condition Report CR-RBS-2008-03661 and determined that suppression pool peak temperature response was acceptable.
The 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality of the suppression pool. The finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee initiated a corrective action program action to re-evaluate long-term suppression pool peak temperature performance but closed the action without its completion [P.1 (d)].
Inspection Report# : 2008006 (pdf)
 
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Testing Programs for 4-kV Circuit Breakers, Class 1E Molded Case Circuit Breakers, and the Emergency Diesel Generators The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, with three examples. Specifically, the team identified that the licensee failed to develop and implement adequate testing programs for 4-kV circuit breakers, Class 1E molded-case circuit breakers, and the emergency diesel generators that met design or vendor requirements and recommendations. In response, the licensee entered these examples in the corrective action program as Condition Reports CR RBS-2008-04379, CR-RBS-2008-3634, CR-RBS-2008-3676 and CR-RBS-2008-3701 and determined there was no loss of safety function for the affected components.
The examples associated with this finding were more than minor because they were associated with the equipment 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined each example was of very low safety significance (Green) because it did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather.
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Implementation of Temporary Installation Procedure The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for five examples of failure to follow the requirements of ADM-0073 Temporary Installation Guidelines" during the installation of modifications to the plant.
Specifically, four modifications were installed in the plant that did not meet the criteria of a temporary installation and one was not removed when no longer needed, as required by the procedure. After identification, the licensee entered the issue into the corrective action program as CR-RBS-2008-3410.
Although the team considered each of the above examples minor in significance, the team determined that this finding, which was associated with design control attribute of the Mitigating Systems cornerstone, was more than minor per Manual Chapter 612, Appendix E, Examples of Minor Issues, Example 4a. The finding involved multiple examples of failure to follow licensee procedural requirements and if left uncorrected it could result in design modifications to the plant that were not properly evaluated, controlled, documented and installed. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect associated with resources in the human performance area because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, those necessary for maintaining long term plant safety by maintenance of design margins, minimization of long-standing
 
equipment issues, minimizing preventative maintenance deferrals, and ensuring maintenance and engineering backlogs which were low enough to support safety [H.2 (a)].
Inspection Report# : 2008006 (pdf)
Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Implementation of Operability Determination Procedure The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the failure to follow procedures to evaluate conditions adverse to quality for impacts on the operability of safety-related equipment. Specifically, the licensee did not assess the impact on operability of previous steam leaks and motor-stall events on the corrosion of magnesium-rotors in safety-related motor-operated valves. The licensee entered this issue into the corrective action program as Condition Reports CR-RBS-2008-3713 and CR-RBS-2008-3766.
The finding was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of safety-related motor-operated valves to respond to initiating events to prevent undesirable consequences. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The cause of the finding had crosscutting aspects associated with the corrective action program in the problem identification and resolution area because the licensee did not thoroughly evaluate the problems with magnesium-rotor corrosion including the extent of the condition and operability impact [P.1(c)].
Inspection Report# : 2008006 (pdf)
Significance:      Jun 28, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Removing Control Building Chilled Water System from Service The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) when operators failed to perform an adequate risk assessment while the Division 1 control building chilled water and control building air conditioning systems were unavailable. Specifically, the inspectors identified that licensee personnel nonconservatively evaluated the on-line risk as Green instead of Yellow. This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis evaluation. The licensee entered this issue into their corrective action program as Condition Report RBS-2008-2687.
Using NRC Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions that put the plant in a higher risk category. Using Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected period is less than 1.0E-6. This finding has a crosscutting aspect in the area of problem identification and resolution component of operating experience because Entergy did not systematically communicate to affected internal stakeholders in a timely manner relevant internal operating experience [P.2(a)].
Inspection Report# : 2008003 (pdf)
 
Barrier Integrity Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate/Untimely Corrective Action for Failure of Magnesium-Rotor Motor-Operated Valves The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, "Corrective Action," for failure to promptly identify magnesium-rotor motor-operated valve degradation. Specifically, the licensee did not identify magnesium-rotor degradation in May 2007 after failure of Valve B21-MOV-FO65A, Reactor Inlet Heater A Outboard Motor Operated Isolation Valve, until after failure of Valve B21-MOV-FO98C, Main Steam Shutoff Valve, in September 2007. The licensee entered this issue into the corrective action program as Condition Reports CR-RBS-2008-3713 and CR-RBS-2008-3766.
This finding was more than minor because Valve B21-MOV-FO98C was associated with the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that the physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. Inspection Manual chapter 0609 Appendix H, "Containment Integrity Significance Determination Process," Table 4.1, indicated that the Main Steam Shutoff Valves do not impact large early release frequency. Based on the results of the Appendix H analysis, the finding was determined to have very low safety significance. This finding had cross-cutting aspects associated with decision-making in the human performance area in that the licensee did not use conservative assumptions in decision-making regarding the likelihood of magnesium-rotor degradation in motor-operated valves [H.1 (b)].
Inspection Report# : 2008006 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Jun 28, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow a Radiation Protection Procedure The inspector reviewed a self-revealing noncited violation of Technical Specification 5.4.1 which resulted from workers failing to follow a radiation protection procedure. On January 13, 2008, three workers attempted to exit the radiologically controlled area and alarmed the personnel contamination monitors. The workers were removing tubes from the Water Box B. The licensee determined radiation protection staff did not follow the radiation work permit planning procedure to use representative radiological surveys for the work performed. The radiation work permit planning did not include previous water box internal and other related surveys which would correspond to the removal of the water box tubes. The licensees investigation found that the contamination levels on the tubes were as high as 150,000 disintegrations per minute per 100 cm2. The licensee revised the radiation work permit to include the actual working conditions and appropriate personnel protective equipment.
Workers failing to follow a radiation protection procedure is a performance deficiency. The finding is greater than minor because, if left uncorrected, the deficiency would become a more radiologically significant safety concern resulting in additional workers unplanned, unintended dose as work continued to be performed under an inadequate radiation work permit. Since this issue involved workers unplanned and unintended dose, the Occupational Radiation
 
Safety Significance Determination Process was used to determine the safety significance. The inspector determined the finding had very low safety significance because: (1) it was not an ALARA finding, (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 control, because the radiation protection staff did not plan work activities consistent with radiological safety by incorporating risk insights and job site conditions of the actual work to be performed during the radiation work permit planning [H.3 (a)].
Inspection Report# : 2008003 (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
 
River Bend 1 2Q/2009 Plant Inspection Findings Initiating Events Significance:        Sep 27, 2008 Identified By: Self-Revealing Item Type: FIN Finding Turbine Building Siding Failure Below Design Specifications A self-revealing finding was identified for wind induced turbine building siding failure that occurred significantly below design specified stress levels as a result of design and installation deficiencies. This resulted in a forced outage to repair transformer damage and to repair the turbine building siding. The licensee missed prior opportunities to identify turbine building siding design and installation deficiencies following damaging wind events in 1992 and 2005. The licensee entered this issue into the corrective action program as Condition Report CR-RBS-2008-5176.
This finding is more than minor because it is associated with the protection against external factors attribute (wind and grid stability) of the initiating events cornerstone and affected 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.
The inspectors evaluated the significance of this finding using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined it to be of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.
Inspection Report# : 2008004 (pdf)
Mitigating Systems Significance:        May 15, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Determinations for a Degraded Diesel Exhaust Pipe The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for twice failing to perform an adequate operability evaluation on the Division II diesel generator after the number 8 cylinder exhaust pipe cracked and later when two of four exhaust flange bolts failed.
The finding is more than minor because it affects the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences.
The team determined that a Phase 3 significance determination was required because the finding screened as potentially risk significant due to potential loss of safety function of a single train. Region IV senior risk analysts performed a Phase 3 significance determination and determined that the issue represents a finding of very low safety significance (Green). This violation has a crosscutting 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 address causes and extent of conditions, as necessary. Specifically the licensee failed to properly prioritize and evaluate for operability a degraded Division II diesel generator Number 8 cylinder exhaust pipe and flange [P.1 (c)].
Inspection Report# : 2009008 (pdf)
 
Significance:      Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment While the Control Building Chilled Water System was Removed from Service The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) involving the failure of operators to perform an adequate risk assessment while the Division 1 control building chilled water was unavailable. Specifically, the inspectors identified that licensee personnel non-conservatively evaluated the on-line risk as Green instead of Yellow. This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis evaluation. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-0862.
Using Inspection Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions regarding the unavailability of mitigating systems that put the plant in a higher risk category. Using Inspection Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component for work practices because Entergy personnel did not effectively follow procedures [H.4(b)].
Inspection Report# : 2009002 (pdf)
Significance:      Mar 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition adverse to quality The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to implement required actions to ensure that conditions were promptly corrected. Specifically, on February 10, 2009, during a review of corrective action documents, the inspectors noted that corrective actions for condition report CR-RBS-2007-03034 were inadequate to correct a condition in which an instrument was not treated as measuring and test equipment. The team noted that corrective action was proposed, but not implemented, and the condition report was closed. The condition which prompted the condition report still existed at the time of the inspection. The licensee entered this issue into corrective action program as condition report CR-RBS-2009-00747.
The failure to implement timely corrective action is a performance deficiency. The finding is greater than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern, such as an improperly calibrated main steam line monitor. The performance deficiency affected the barrier integrity cornerstone in that the proper calibration of the main steam line monitors is necessary to ensure proper isolation of containment in the event of fuel damage. Using Phase 1 worksheet from Manual Chapter 0609, Significance Determination Process, this finding was determined to have very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system; did not represent an actual open pathway in the physical integrity of the reactor containment and heat removal components, 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 did not use conservative assumptions to demonstrate that the decision to close the condition report with no further action was appropriate (H1.b).
Inspection Report# : 2009006 (pdf)
Significance:      Sep 27, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Staging the Station Blackout Diesel Generator during Severe Weather The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving the failure to have an adequate procedure to ensure the availability of on site emergency ac power sources following the four-hour coping period of a postulated station blackout. Specifically, station procedures did not ensure that the station blackout diesel
 
generator would be reliably deployed to fulfill its intended function during sustained high winds. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-5050.
This finding is more than minor because it is associated with the protection against external factors attribute (wind and grid stability) 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 inspectors evaluated the significance of this finding using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined it to be of very low safety significance because it did not result in an actual loss of safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 27, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Corrective Actions Results in Multiple Failures of Standby Service Water Switchgear Room Ventilation Fans A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to take adequate corrective actions in response to a condition adverse to quality resulting in repetitive failures of the standby service water switchgear room ventilation fans. Following failure of the switchgear fans in July 2008, the licensee found that inappropriate flow switch settings on the fans had been identified in a condition report in October 1999, but no actions had been taken to correct the condition. Subsequently, more failures of the standby service water switchgear room ventilation fans occurred, including nineteen in the past three and one half years, many of which were attributed to flow switch issues. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-5761.
The finding was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone, and it directly affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to preclude 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 the condition did not result in an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. This finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee failed to maintain long term plant safety by minimization of long standing equipment issues [H.2(a)].
Inspection Report# : 2008004 (pdf)
Significance:        Sep 27, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Transformer Yard Maintenance While Shut Down The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the licensees failure to assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, while conducting maintenance in the transformer yard during severe weather with high pressure core spray inoperable, the licensee did not assess the affects on the shutdown risk. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-05383.
The inspectors determined this finding was more than minor since it was similar to Manual Chapter 0612, Appendix E, Example 7.e, and since it caused the licensees risk model to change from a Green to Yellow risk window. In accordance with NRC Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management, the inspectors requested that a senior reactor analyst evaluate the risk of this condition. The analyst determined that this finding was of very low risk significance because the associated risk deficit was less than 1.0E-6.
 
Inspection Report# : 2008004 (pdf)
Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Eight Examples of a Failure to Meet 10 CFR Part 50, Appendix B, "Design Control" The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with eight examples.
* Example 1: Non-conservative inputs and assumptions used without adequate technical justification to evaluate the minimum terminal voltage and actuator output torque for safety-related motor operated valves. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03339.
* Example 2: Failure to perform a conservative analysis to ensure that Technical Specification Setpoints were adequate. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03911.
* Example 3: Non-conservative inputs and methodologies used in calculating control circuit voltages to safety-related 480V motor operated valves motor-operated valve and motors that would be required to operate for mitigation of design bases events. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03858.
* Example 4: Failure to evaluate E12-MOV-F042A, residual heat removal injection valve, and E12-MOV-F064A, residual heat removal minimum flow valve, to verify adequate voltage would be available to operate the associated 120VAC control circuit devices. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03641
* Example 5: Inadequate design basis documentation for hydrogen concentration control in the Division I and II Battery Rooms in the control building. After identification, the licensee entered the issue into the corrective action program as Condition Reports CR-RBS-2008-02566 and CR-RBS-2008-03403.
* Example 6: Failure to ensure design basis information for safety related 125VDC batteries was controlled and correctly translated into procedures and instructions. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03659.
* Example 7: Failure to maintain adequate design basis calculations for ultimate heat sink loading. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-3712.
* Example 8: Failure to account for the technical specification allowed emergency diesel generator frequency variation in the diesel loading calculation. After identification, the licensee entered the issue into the corrective action program as Condition Report CR-RBS-2008-03556.
The examples associated with this finding were more than minor per Manual Chapter 612, Appendix E, Appendix E, Examples of Minor Issues, Example 3j, in that each example resulted in a condition where there was reasonable doubt on the operability of a system or component. The finding 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined each example was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality.
Inspection Report# : 2008006 (pdf)
 
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Recalculate Suppression Pool Peak Temperature Rseponse The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that, design control measures for verifying the adequacy of design were not implemented. Specifically, the licensee did not recalculate suppression pool peak temperature response when a more severe single failure condition was identified. In response, the licensee entered this issue in the corrective action program as Condition Report CR-RBS-2008-03661 and determined that suppression pool peak temperature response was acceptable.
The 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality of the suppression pool. The finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee initiated a corrective action program action to re-evaluate long-term suppression pool peak temperature performance but closed the action without its completion [P.1 (d)].
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Testing Programs for 4-kV Circuit Breakers, Class 1E Molded Case Circuit Breakers, and the Emergency Diesel Generators The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, with three examples. Specifically, the team identified that the licensee failed to develop and implement adequate testing programs for 4-kV circuit breakers, Class 1E molded-case circuit breakers, and the emergency diesel generators that met design or vendor requirements and recommendations. In response, the licensee entered these examples in the corrective action program as Condition Reports CR RBS-2008-04379, CR-RBS-2008-3634, CR-RBS-2008-3676 and CR-RBS-2008-3701 and determined there was no loss of safety function for the affected components.
The examples associated with this finding were more than minor because they were associated with the equipment 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. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined each example was of very low safety significance (Green) because it did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather.
Inspection Report# : 2008006 (pdf)
Significance:        Aug 26, 2008
 
Identified By: NRC Item Type: NCV NonCited Violation Inadequate Implementation of Temporary Installation Procedure The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for five examples of failure to follow the requirements of ADM-0073 Temporary Installation Guidelines" during the installation of modifications to the plant.
Specifically, four modifications were installed in the plant that did not meet the criteria of a temporary installation and one was not removed when no longer needed, as required by the procedure. After identification, the licensee entered the issue into the corrective action program as CR-RBS-2008-3410.
Although the team considered each of the above examples minor in significance, the team determined that this finding, which was associated with design control attribute of the Mitigating Systems cornerstone, was more than minor per Manual Chapter 612, Appendix E, Examples of Minor Issues, Example 4a. The finding involved multiple examples of failure to follow licensee procedural requirements and if left uncorrected it could result in design modifications to the plant that were not properly evaluated, controlled, documented and installed. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect associated with resources in the human performance area because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, those necessary for maintaining long term plant safety by maintenance of design margins, minimization of long-standing equipment issues, minimizing preventative maintenance deferrals, and ensuring maintenance and engineering backlogs which were low enough to support safety [H.2 (a)].
Inspection Report# : 2008006 (pdf)
Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Implementation of Operability Determination Procedure The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the failure to follow procedures to evaluate conditions adverse to quality for impacts on the operability of safety-related equipment. Specifically, the licensee did not assess the impact on operability of previous steam leaks and motor-stall events on the corrosion of magnesium-rotors in safety-related motor-operated valves. The licensee entered this issue into the corrective action program as Condition Reports CR-RBS-2008-3713 and CR-RBS-2008-3766.
The finding was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of safety-related motor-operated valves to respond to initiating events to prevent undesirable consequences. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, a Phase 1 screening was performed and determined the finding was of very low safety significance (Green) because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The cause of the finding had crosscutting aspects associated with the corrective action program in the problem identification and resolution area because the licensee did not thoroughly evaluate the problems with magnesium-rotor corrosion including the extent of the condition and
 
operability impact [P.1(c)].
Inspection Report# : 2008006 (pdf)
Barrier Integrity Significance:      Aug 26, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate/Untimely Corrective Action for Failure of Magnesium-Rotor Motor-Operated Valves The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, "Corrective Action," for failure to promptly identify magnesium-rotor motor-operated valve degradation. Specifically, the licensee did not identify magnesium-rotor degradation in May 2007 after failure of Valve B21-MOV-FO65A, Reactor Inlet Heater A Outboard Motor Operated Isolation Valve, until after failure of Valve B21-MOV-FO98C, Main Steam Shutoff Valve, in September 2007. The licensee entered this issue into the corrective action program as Condition Reports CR-RBS-2008-3713 and CR-RBS-2008-3766.
This finding was more than minor because Valve B21-MOV-FO98C was associated with the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that the physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRCs regulatory function, and was not the result of any willful violation of NRC requirements. Inspection Manual chapter 0609 Appendix H, "Containment Integrity Significance Determination Process," Table 4.1, indicated that the Main Steam Shutoff Valves do not impact large early release frequency. Based on the results of the Appendix H analysis, the finding was determined to have very low safety significance. This finding had cross-cutting aspects associated with decision-making in the human performance area in that the licensee did not use conservative assumptions in decision-making regarding the likelihood of magnesium-rotor degradation in motor-operated valves [H.1 (b)].
Inspection Report# : 2008006 (pdf)
Emergency Preparedness Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Formally Critique an Emergency Plan Weakness The inspectors identified a violation of 10 CFR 50.47(b)(14) for failure to identify and critique a nonrisk significant planning standard weakness demonstrated during a site emergency preparedness drill. Specifically, the licensee demonstrated a weakness in controlling radiological exposures for emergency workers during an emergency, without key emergency response organization decision maker consideration or input, when simulated emergency workers were left in containment during changing radiological conditions. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-02458.
This finding is more than minor because it is associated with the emergency response organization performance attribute of the Emergency Preparedness Cornerstone which ensures the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors evaluated the significance of this finding using Sheet 1, Failure to Comply, of Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and determined it to be of very low safety significance (Green) because the finding was a failure to comply with the requirements of 10 CFR 50.47(b)
 
(14), the finding was associated with an emergency preparedness planning standard, the associated planning standard was not risk significant as defined by Manual Chapter 0609, Appendix B, and the finding was not a functional failure of the planning standard function. The inspectors determined that the finding has a crosscutting aspect in the area of problem identification and resolution because the licensee did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1(a)].
Inspection Report# : 2009003 (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 Last modified : August 31, 2009
 
River Bend 1 3Q/2009 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        May 15, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Determinations for a Degraded Diesel Exhaust Pipe The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for twice failing to perform an adequate operability evaluation on the Division II diesel generator after the number 8 cylinder exhaust pipe cracked and later when two of four exhaust flange bolts failed.
The finding is more than minor because it affects the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences.
The team determined that a Phase 3 significance determination was required because the finding screened as potentially risk significant due to potential loss of safety function of a single train. Region IV senior risk analysts performed a Phase 3 significance determination and determined that the issue represents a finding of very low safety significance (Green). This violation has a crosscutting 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 address causes and extent of conditions, as necessary. Specifically the licensee failed to properly prioritize and evaluate for operability a degraded Division II diesel generator Number 8 cylinder exhaust pipe and flange [P.1 (c)].
Inspection Report# : 2009008 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment While the Control Building Chilled Water System was Removed from Service The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) involving the failure of operators to perform an adequate risk assessment while the Division 1 control building chilled water was unavailable. Specifically, the inspectors identified that licensee personnel non-conservatively evaluated the on-line risk as Green instead of Yellow. This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis evaluation. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-0862.
Using Inspection Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions regarding the unavailability of mitigating systems that put the plant in a higher risk category. Using Inspection Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component for work practices because Entergy personnel did not effectively follow procedures [H.4(b)].
Inspection Report# : 2009002 (pdf)
Significance:        Mar 30, 2009
 
Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition adverse to quality The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to implement required actions to ensure that conditions were promptly corrected. Specifically, on February 10, 2009, during a review of corrective action documents, the inspectors noted that corrective actions for condition report CR-RBS-2007-03034 were inadequate to correct a condition in which an instrument was not treated as measuring and test equipment. The team noted that corrective action was proposed, but not implemented, and the condition report was closed. The condition which prompted the condition report still existed at the time of the inspection. The licensee entered this issue into corrective action program as condition report CR-RBS-2009-00747.
The failure to implement timely corrective action is a performance deficiency. The finding is greater than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern, such as an improperly calibrated main steam line monitor. The performance deficiency affected the barrier integrity cornerstone in that the proper calibration of the main steam line monitors is necessary to ensure proper isolation of containment in the event of fuel damage. Using Phase 1 worksheet from Manual Chapter 0609, Significance Determination Process, this finding was determined to have very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system; did not represent an actual open pathway in the physical integrity of the reactor containment and heat removal components, 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 did not use conservative assumptions to demonstrate that the decision to close the condition report with no further action was appropriate (H1.b).
Inspection Report# : 2009006 (pdf)
Barrier Integrity Emergency Preparedness Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Formally Critique an Emergency Plan Weakness The inspectors identified a violation of 10 CFR 50.47(b)(14) for failure to identify and critique a nonrisk significant planning standard weakness demonstrated during a site emergency preparedness drill. Specifically, the licensee demonstrated a weakness in controlling radiological exposures for emergency workers during an emergency, without key emergency response organization decision maker consideration or input, when simulated emergency workers were left in containment during changing radiological conditions. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-02458.
This finding is more than minor because it is associated with the emergency response organization performance attribute of the Emergency Preparedness Cornerstone which ensures the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors evaluated the significance of this finding using Sheet 1, Failure to Comply, of Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and determined it to be of very low safety significance (Green) because the finding was a failure to comply with the requirements of 10 CFR 50.47(b)
(14), the finding was associated with an emergency preparedness planning standard, the associated planning standard was not risk significant as defined by Manual Chapter 0609, Appendix B, and the finding was not a functional failure of the planning standard function. The inspectors determined that the finding has a crosscutting aspect in the area of problem identification and resolution because the licensee did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1(a)].
 
Inspection Report# : 2009003 (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 Last modified : December 10, 2009
 
River Bend 1 4Q/2009 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Scaffold Construction The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for the failure of maintenance personnel to control scaffold erection per procedure. This failure resulted in the licensee installing 31 scaffolds in safety related areas that required either rework or an engineering evaluation to resolve as built deviations from the minimum seismic separation requirements. As a result, the design function of the safety related equipment was potentially adversely affected. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-3963.
The failure to erect scaffolds in accordance with procedures is a performance deficiency. This finding is more than minor because it is similar to Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Section 4, Example a, because Entergy had routinely failed to perform the requisite engineering evaluation and because it was associated with the protection against external events 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 in order to prevent undesirable consequences. The finding was determined to be of very low risk significance (Green) because no actual loss of safety function occurred and the finding did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the area of human performance, work practices, because the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Reactor Core Isolation Cooling System Seismic Design The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to implement measures to ensure that the seismic design basis for the reactor core isolation cooling turbine governor hydraulic system was correctly translated into the specifications, drawings, procedures, or instructions. This resulted in work to reroute the piping and an engineering evaluation to resolve seismic concerns. The licensee entered this issue into their corrective action program as Condition Report CR RBS 2009 3747.
The failure to implement design control features for the seismic design of the reactor core isolation cooling system is a performance deficiency. This finding was more than minor because it is similar to Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Section 5, Example a, in that the reactor core isolation cooling turbine was returned to service without the seismic spacing required by the original design or completion of an evaluation for the as left condition. This resulted in rework and additional engineering analysis to correctly resolve the seismic qualification concerns. The performance deficiency also affected the mitigating systems cornerstone attribute of external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Attachment A, Phase 1 - Initial Screening and Characterization of Findings,
 
for the mitigating systems cornerstone. After answering no to all five questions in the mitigating systems cornerstone column of Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance. This finding does not have a crosscutting aspect because the performance deficiency occurred in 1989 and is not reflective of current plant performance.
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Standby Liquid Control System Test Tank Remained Drained The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for the failure of operations personnel to provide adequate procedural guidance to preclude water intrusion into the nonseismically qualified standby liquid control system test tank which resulted in the degradation of both trains of the standby liquid control system. The licensee entered this issue into their corrective action program as Condition Report CR RBS 2009 3862.
The failure to provide appropriate procedures to keep the standby liquid control test tank drained is a performance deficiency. The finding is more than minor because it affects the protection against external events attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance because the finding was not a design or qualification deficiency, did not represent a loss of a system/train safety function, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the area of problem identification and resolutions corrective action program because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, the licensee failed to address the cause of inadvertent water intrusion into the standby liquid control test tank in a timely manner to prevent the common mode failure of both trains of standby liquid control [P.1(d)].
Inspection Report# : 2009004 (pdf)
Significance:        May 15, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Determinations for a Degraded Diesel Exhaust Pipe The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for twice failing to perform an adequate operability evaluation on the Division II diesel generator after the number 8 cylinder exhaust pipe cracked and later when two of four exhaust flange bolts failed.
The finding is more than minor because it affects the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences.
The team determined that a Phase 3 significance determination was required because the finding screened as potentially risk significant due to potential loss of safety function of a single train. Region IV senior risk analysts performed a Phase 3 significance determination and determined that the issue represents a finding of very low safety significance (Green). This violation has a crosscutting 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 address causes and extent of conditions, as necessary. Specifically the licensee failed to properly prioritize and evaluate for operability a degraded Division II diesel generator Number 8 cylinder exhaust pipe and flange [P.1 (c)].
Inspection Report# : 2009008 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment While the Control Building Chilled Water System was Removed from Service The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4) involving the failure of operators to
 
perform an adequate risk assessment while the Division 1 control building chilled water was unavailable. Specifically, the inspectors identified that licensee personnel non-conservatively evaluated the on-line risk as Green instead of Yellow. This resulted in an unrecognized increase in the level of risk as determined by Entergys probabilistic safety analysis evaluation. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-0862.
Using Inspection Manual Chapter 0612, Appendix E, Section 3, Item 7(e), the finding is more than minor because the licensees risk assessment had errors and incorrect assumptions regarding the unavailability of mitigating systems that put the plant in a higher risk category. Using Inspection Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance because the incremental core damage probability deficit for the affected time period is less than 1.0E-6. This finding has a crosscutting aspect in the area of human performance component for work practices because Entergy personnel did not effectively follow procedures [H.4(b)].
Inspection Report# : 2009002 (pdf)
Barrier Integrity Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Containment Closure Procedure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of operations personnel to fully implement a station procedure to control obstructions in primary containment openings in Modes 4 and 5. The failure to follow procedure challenged the licensees ability to establish containment closure. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-4296.
The failure to implement the containment closure procedure is a performance deficiency. This finding is more than minor because it affected the configuration control attribute of the barrier integrity objective to provide reasonable assurance that the physical design barriers (containment) will protect the public from radionuclide releases. Using Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, the finding was assessed as a Type B finding because it is related to a degraded condition that has potentially important implications for the integrity of the containment without affecting the likelihood of core damage and was of very low significance because the licensee did not lose the capability to close containment when planned. The finding has a crosscutting aspect in the area of human performance, work control, because the licensee failed to appropriately coordinate work activities (identifying cables, quick disconnects, removing unidentified cables) to address the operational impact of those work activities on containment operability [H.3(b)].
Inspection Report# : 2009004 (pdf)
Significance:        Mar 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition adverse to quality The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to implement required actions to ensure that conditions were promptly corrected. Specifically, on February 10, 2009, during a review of corrective action documents, the inspectors noted that corrective actions for condition report CR-RBS-2007-03034 were inadequate to correct a condition in which an instrument was not treated as measuring and test equipment. The team noted that corrective action was proposed, but not implemented, and the condition report was closed. The condition which prompted the condition report still existed at the time of the inspection. The licensee entered this issue into corrective action program as condition report CR-RBS-2009-00747.
The failure to implement timely corrective action is a performance deficiency. The finding is greater than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety
 
concern, such as an improperly calibrated main steam line monitor. The performance deficiency affected the barrier integrity cornerstone in that the proper calibration of the main steam line monitors is necessary to ensure proper isolation of containment in the event of fuel damage. Using Phase 1 worksheet from Manual Chapter 0609, Significance Determination Process, this finding was determined to have very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system; did not represent an actual open pathway in the physical integrity of the reactor containment and heat removal components, 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 did not use conservative assumptions to demonstrate that the decision to close the condition report with no further action was appropriate (H1.b).
Inspection Report# : 2009006 (pdf)
Emergency Preparedness Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Formally Critique an Emergency Plan Weakness The inspectors identified a violation of 10 CFR 50.47(b)(14) for failure to identify and critique a nonrisk significant planning standard weakness demonstrated during a site emergency preparedness drill. Specifically, the licensee demonstrated a weakness in controlling radiological exposures for emergency workers during an emergency, without key emergency response organization decision maker consideration or input, when simulated emergency workers were left in containment during changing radiological conditions. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-02458.
This finding is more than minor because it is associated with the emergency response organization performance attribute of the Emergency Preparedness Cornerstone which ensures the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors evaluated the significance of this finding using Sheet 1, Failure to Comply, of Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and determined it to be of very low safety significance (Green) because the finding was a failure to comply with the requirements of 10 CFR 50.47(b)
(14), the finding was associated with an emergency preparedness planning standard, the associated planning standard was not risk significant as defined by Manual Chapter 0609, Appendix B, and the finding was not a functional failure of the planning standard function. The inspectors determined that the finding has a crosscutting aspect in the area of problem identification and resolution because the licensee did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1(a)].
Inspection Report# : 2009003 (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 Last modified : March 01, 2010
 
River Bend 1 1Q/2010 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Main Steam Line Plug Seal Failure Results in Loss of Reactor Cavity Inventory A self-revealing noncited violation of 10CFR50 Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the failure to follow the procedure for making a permanent plant modification and provide adequate procedures for installation and use of the main steam line plugs following a main steam line plug design change. This failure resulted in draining approximately 5,000 gallons of water from the upper reactor cavity pool to the drywell and a manual actuation of low pressure coolant injection to restore cavity pool water level. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-4681.
The finding was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issues Screening, because the finding affected the initiating events cornerstone attribute of configuration control and the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the finding in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors reviewed Section II.B.(1) of Checklist 7 and determined that the finding required a Phase 2 analysis because the finding involved procedures that affected steam line plug seal configuration and resulted in inventory loss from the upper reactor cavity pool. The senior reactor analyst determined that, because of the special circumstances of this event, the use of a qualitative assessment using Inspection Manual Chapter 0609, Appendix M, was more appropriate than the risk tools provided in Inspection Manual Chapter 0609, Appendix G. This is because the draindown event was self-limiting, such that the inventory excursion could not have drained reactor cavity level below the level of the main steam lines, and that even with the failure of operators to take actions, the core would have remained covered with no challenges to the shutdown cooling system. Therefore, the event in the worst case would have been transparent to the core. Also, the displaced inventory posed no threat to any of the plant's mitigating systems. The inspectors concluded that the finding was of very low safety significance (Green). There is no crosscutting aspect associated with this violation because the finding does not reflect current licensee performance.
Inspection Report# : 2009005 (pdf)
Mitigating Systems Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Evaluate a Revised Equipment Tag Out A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for defeating the Division I emergency systems automatic start functions caused by the failure to follow a work implementation and closeout procedure when changing the work scope and tag out boundaries for a safety-related maintenance activity. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-06151.
The finding was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the finding did not represent an actual loss of safety function. This finding has a crosscutting aspect in the area of
 
human performance, work control because the licensee did not appropriately plan activities by incorporating actions to address operational impact and risk for the work scope changes [H.3(a)].
Inspection Report# : 2010002 (pdf)
Significance:      Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Control Building Chiller Operability During Low Service Water Temperatures The inspectors identified a Green noncited violation of Technical Specification 3.7.3 for exceeding the control room air conditioning system thirty day allowed outage time for one inoperable subsystem and the seven day allowed outage time for two inoperable subsystems and failing to enter Modes 3 and 4, as specified. Specifically, during accident conditions the control building chillers were not able to remove the design basis heat load while operating with low standby cooling water temperatures. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-01593, CR-RBS-2010-01817 and CR-RBS-2010-01667.
The performance deficiency was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issue Disposition Screening, because the finding was associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability of multiple safety-related systems and components to 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 because the finding resulted in an actual loss of safety function of a single train for greater than its technical specification allowed outage time and required a Phase 2 analysis. However, the Phase 2 presolved table and worksheets did not contain appropriate target sets to estimate accurately the risk impact of the finding. Therefore, the senior reactor analyst performed a Phase 3 analysis. The estimated change in core damage frequency was 2.3E-8/yr. Therefore, the inspectors determined the significance of the finding was Green. This finding was not assigned a crosscutting aspect because it does not reflect current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:      Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Adequately Verify a Suitable Replacement Part Essential for Emergency Diesel Generator Operation A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for the failure to adequately verify suitable replacement parts essential to the operation of emergency diesel generator Division I. This resulted in multiple intercooler flange bolts failing from low stress, high cycle fatigue. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-06148.
The finding is also similar to example 3j of Manual Chapter 0612 Appendix E. Specifically, the number of bolting failures placed the emergency diesel generators operability in doubt and an engineering analysis had to be performed to prove operability. In accordance with Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the inspectors performed a significant determination process Phase 1 screening and determined that the finding was of very low safety significance (Green) because a licensee analysis concluded that the bolts that were projected to fail during the emergency diesel generator mission time of thirty days would not result in an actual loss of system safety function. The inspectors determined that the finding had a crosscutting aspect in the area of human performance resources in that the licensee failed to ensure that equipment was adequate for maintaining long term plant safety by maintenance of design margins [H.2(a)].
Inspection Report# : 2010002 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Scaffold Construction The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for the failure of maintenance
 
personnel to control scaffold erection per procedure. This failure resulted in the licensee installing 31 scaffolds in safety related areas that required either rework or an engineering evaluation to resolve as built deviations from the minimum seismic separation requirements. As a result, the design function of the safety related equipment was potentially adversely affected. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-3963.
The failure to erect scaffolds in accordance with procedures is a performance deficiency. This finding is more than minor because it is similar to Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Section 4, Example a, because Entergy had routinely failed to perform the requisite engineering evaluation and because it was associated with the protection against external events 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 in order to prevent undesirable consequences. The finding was determined to be of very low risk significance (Green) because no actual loss of safety function occurred and the finding did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the area of human performance, work practices, because the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Reactor Core Isolation Cooling System Seismic Design The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to implement measures to ensure that the seismic design basis for the reactor core isolation cooling turbine governor hydraulic system was correctly translated into the specifications, drawings, procedures, or instructions. This resulted in work to reroute the piping and an engineering evaluation to resolve seismic concerns. The licensee entered this issue into their corrective action program as Condition Report CR RBS 2009 3747.
The failure to implement design control features for the seismic design of the reactor core isolation cooling system is a performance deficiency. This finding was more than minor because it is similar to Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Section 5, Example a, in that the reactor core isolation cooling turbine was returned to service without the seismic spacing required by the original design or completion of an evaluation for the as left condition. This resulted in rework and additional engineering analysis to correctly resolve the seismic qualification concerns. The performance deficiency also affected the mitigating systems cornerstone attribute of external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Attachment A, Phase 1 - Initial Screening and Characterization of Findings, for the mitigating systems cornerstone. After answering no to all five questions in the mitigating systems cornerstone column of Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance. This finding does not have a crosscutting aspect because the performance deficiency occurred in 1989 and is not reflective of current plant performance.
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Standby Liquid Control System Test Tank Remained Drained The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for the failure of operations personnel to provide adequate procedural guidance to preclude water intrusion into the nonseismically qualified standby liquid control system test tank which resulted in the degradation of both trains of the standby liquid control system. The licensee entered this issue into their corrective action program as Condition Report CR RBS 2009 3862.
The failure to provide appropriate procedures to keep the standby liquid control test tank drained is a performance deficiency. The finding is more than minor because it affects the protection against external events attribute of the
 
mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance because the finding was not a design or qualification deficiency, did not represent a loss of a system/train safety function, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the area of problem identification and resolutions corrective action program because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, the licensee failed to address the cause of inadvertent water intrusion into the standby liquid control test tank in a timely manner to prevent the common mode failure of both trains of standby liquid control [P.1(d)].
Inspection Report# : 2009004 (pdf)
Significance:        May 15, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Determinations for a Degraded Diesel Exhaust Pipe The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for twice failing to perform an adequate operability evaluation on the Division II diesel generator after the number 8 cylinder exhaust pipe cracked and later when two of four exhaust flange bolts failed.
The finding is more than minor because it affects the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences.
The team determined that a Phase 3 significance determination was required because the finding screened as potentially risk significant due to potential loss of safety function of a single train. Region IV senior risk analysts performed a Phase 3 significance determination and determined that the issue represents a finding of very low safety significance (Green). This violation has a crosscutting 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 address causes and extent of conditions, as necessary. Specifically the licensee failed to properly prioritize and evaluate for operability a degraded Division II diesel generator Number 8 cylinder exhaust pipe and flange [P.1 (c)].
Inspection Report# : 2009008 (pdf)
Barrier Integrity Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Containment Closure Procedure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of operations personnel to fully implement a station procedure to control obstructions in primary containment openings in Modes 4 and 5. The failure to follow procedure challenged the licensees ability to establish containment closure. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-4296.
The failure to implement the containment closure procedure is a performance deficiency. This finding is more than minor because it affected the configuration control attribute of the barrier integrity objective to provide reasonable assurance that the physical design barriers (containment) will protect the public from radionuclide releases. Using Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, the finding was assessed as a Type B finding because it is related to a degraded condition that has potentially important implications for the integrity of the containment without affecting the likelihood of core damage and was of very low significance because the licensee did not lose the capability to close containment when planned. The finding has a crosscutting aspect in the area of human performance, work control, because the licensee failed to appropriately coordinate work activities (identifying cables, quick disconnects, removing unidentified cables) to address the operational impact of those work activities on containment operability [H.3(b)].
 
Inspection Report# : 2009004 (pdf)
Emergency Preparedness Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Formally Critique an Emergency Plan Weakness The inspectors identified a violation of 10 CFR 50.47(b)(14) for failure to identify and critique a nonrisk significant planning standard weakness demonstrated during a site emergency preparedness drill. Specifically, the licensee demonstrated a weakness in controlling radiological exposures for emergency workers during an emergency, without key emergency response organization decision maker consideration or input, when simulated emergency workers were left in containment during changing radiological conditions. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-02458.
This finding is more than minor because it is associated with the emergency response organization performance attribute of the Emergency Preparedness Cornerstone which ensures the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors evaluated the significance of this finding using Sheet 1, Failure to Comply, of Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and determined it to be of very low safety significance (Green) because the finding was a failure to comply with the requirements of 10 CFR 50.47(b)
(14), the finding was associated with an emergency preparedness planning standard, the associated planning standard was not risk significant as defined by Manual Chapter 0609, Appendix B, and the finding was not a functional failure of the planning standard function. The inspectors determined that the finding has a crosscutting aspect in the area of problem identification and resolution because the licensee did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1(a)].
Inspection Report# : 2009003 (pdf)
Occupational Radiation Safety Public Radiation Safety Significance:      Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure the Emergency Contact had Knowledge About a Shipment The inspectors identified a noncited violation of 10 CFR 71.5 and 49 CFR Part 172.604(a) for a failure to ensure that the shift manager, whose phone number was listed as the required 24-hour emergency phone number on shipping documents, was knowledgeable about the radioactive waste shipment that left site on December 16, 2009, and had immediate access to a person who had specific information on the shipment. Specifically, the shift manager was listed as the required 24-hour contact; however, the shift managers (on multiple shifts) were not provided with documentation or information about the shipments that left the site on December 16, 2009. Although the shift manager would have eventually contacted a knowledgeable person, this delay would not have resulted in immediate access to the person with information related to the shipment. The licensee immediately provided the shift manager a copy of the shipping documentation, briefed the shift manager, and entered this issue into their corrective action program as Condition Report CR-RBS-2009-06419.
This performance deficiency was more than minor because it adversely affected the public radiation safety
 
cornerstone to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain in that the failure to have shipment information immediately available could restrict the actions of fire department and/or rescue personnel responding to an accident. When processed through the Public Radiation Safety Determination Process, the finding was determined to be of very low safety significance because the finding:
(1) was associated with radioactive material control, (2) involved the licensees program for radioactive material transportation, (3) did not cause radiation limits to be exceeded, (4) did not involve a breach of package during transit, (5) did not involve a certificate of compliance finding, (6) did not involve a low level burial ground nonconformance, and (7) did not involve a failure to make notifications. The inspectors determined the finding had a crosscutting aspect in area of resources, associated with documentation, because the licensees procedures did not provide guidance on informing the control room about shipments and thus, the procedures were not complete, accurate nor up-to-date [H.2 (c)].
Inspection Report# : 2009005 (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 : May 26, 2010
 
River Bend 1 2Q/2010 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Main Steam Line Plug Seal Failure Results in Loss of Reactor Cavity Inventory A self-revealing noncited violation of 10CFR50 Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the failure to follow the procedure for making a permanent plant modification and provide adequate procedures for installation and use of the main steam line plugs following a main steam line plug design change. This failure resulted in draining approximately 5,000 gallons of water from the upper reactor cavity pool to the drywell and a manual actuation of low pressure coolant injection to restore cavity pool water level. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-4681.
The finding was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issues Screening, because the finding affected the initiating events cornerstone attribute of configuration control and the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the finding in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors reviewed Section II.B.(1) of Checklist 7 and determined that the finding required a Phase 2 analysis because the finding involved procedures that affected steam line plug seal configuration and resulted in inventory loss from the upper reactor cavity pool. The senior reactor analyst determined that, because of the special circumstances of this event, the use of a qualitative assessment using Inspection Manual Chapter 0609, Appendix M, was more appropriate than the risk tools provided in Inspection Manual Chapter 0609, Appendix G. This is because the draindown event was self-limiting, such that the inventory excursion could not have drained reactor cavity level below the level of the main steam lines, and that even with the failure of operators to take actions, the core would have remained covered with no challenges to the shutdown cooling system. Therefore, the event in the worst case would have been transparent to the core. Also, the displaced inventory posed no threat to any of the plant's mitigating systems. The inspectors concluded that the finding was of very low safety significance (Green). There is no crosscutting aspect associated with this violation because the finding does not reflect current licensee performance.
Inspection Report# : 2009005 (pdf)
Mitigating Systems Significance:        Jun 02, 2010 Identified By: NRC Item Type: VIO Violation Failure to Ensure at Least One Train of Equipment Necessary to Achieve Hot Shutdown Conditions is Free of Fire Damage The team identified a cited violation of License Condition 2.C.(10), Fire Protection, for failing to ensure that the Division 1 standby service water support system to the Division 1 emergency diesel generator, which was required to achieve safe shutdown, was protected such that it remained free from fire damage under all conditions. This condition was identified by the licensee in May 2007, and entered into their corrective action program as a significant non-conforming condition in CR-RBS-2007-02102. The licensee subsequently initiated compensatory measures in the form of manual actions to protect the Division 1 emergency diesel generator. This issue was documented as a licensee-identified noncited violation in Inspection Report 2009002. River Bend has subsequently completed two refueling outages, six forced outages, and one emergency diesel generator work window of sufficient duration since
 
identification of this condition and failed to correct the non-conformance. The team determined that schedule changes resulted in a new completion date of January 2011.
The failure to ensure that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) was free of fire damage and to correct this significant non-conforming condition in a timely manner is a performance deficiency. This performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A Phase 3 SDP risk assessment was performed by a senior reactor analyst. The bounding change in conditional core damage frequency for a 1-year exposure is the Fire Mitigation Frequency (4.30E-08/year) multiplied by the change in conditional core damage probability (0.9) for a value of 3.87E-08/year. This value indicates the finding has very low safety significance (Green). Because the licensee failed to correct this violation, this violation is being treated as a cited violation, consistent with the NRC Enforcement Policy. This finding had a crosscutting aspect in the Work Control component of the Human Performance area because the licensee did not appropriately plan work activities to support long-term equipment reliability by limiting temporary modifications, operator workarounds, safety systems unavailability, and reliance on manual actions [H.3(b)]. (Section 1R05.01)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Alternative Shutdown Procedure Could be Implemented as Written The team identified a noncited violation of Technical Specification 5.4.1.d, Fire Protection Program Implementation. Specifically, Procedure AOP-0031 Shutdown from Outside the Main Control Room, Revision 307, had steps that could not be implemented as written. Two steps were to be performed before the necessary ac power was available, and two steps required diagnostic assessment without the availability of instrumentation.
The failure to ensure that Procedure AOP-0031, Revision 307 could be implemented as written is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Attachment 2 to Appendix F, Fire Protection Significance Determination Process, this issue was determined to be a safe shutdown finding, and was assigned a degradation rating of Low because the examples involved procedural deficiencies that could be compensated for by operator experience. Since this finding was assigned a low degradation rating, the safety significance screened as very low (Green). This finding was entered into the licensees corrective action program as CR-RBS-2010-01592, CR-RBS-2010-01831, CR-RBS-2010-01775, CR-RBS-2010-01821, and CR-RBS-2010-1846. This finding had a crosscutting aspect in the Resources component of the Human Performance area, in that the licensee did not ensure that procedures were complete, accurate, and up to date to assure nuclear safety [H.2.(c)]. (Section 1R05.05.b.1)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement and Maintain in Effect all Provisions of the Approved Fire Protection Program The team identified a noncited violation of License Condition 2.C.(10), Fire Protection, for the failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the team identified, during a timed walkdown of the procedure that it took operators over 6 minutes to isolate feedwater, but the simulator showed that the steam lines could be flooded in 2 minutes. Overfilling the reactor pressure vessel and flooding the main steam lines could make reactor core isolation cooling unavailable. Reactor core isolation cooling was credited
 
for decay heat removal and inventory control in the event of a fire.
The failure to ensure that feedwater would be isolated prior to overfilling the reactor pressure vessel and flooding the main steam lines making reactor core isolation cooling unavailable is a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a control room fire that led to control room abandonment. The Phase 3 evaluation determined that the finding had very low safety significance because a fire in only one of 109 electrical cabinets in the control room could result in this overfill event. The finding was entered into the licensees corrective action program as CR-RBS-2010-01808. The finding did not have a crosscutting aspect since it was not indicative of current performance, in that the licensee had established the incorrect response time more than three years prior to this finding. (Section 1R05.05.b.2)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement and Maintain in Effect all Provisions of the Approved Fire Protection Program The team identified a noncited violation of License Condition 2.C.(10), Fire Protection, related to the licensee's failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, during testing required by the approved fire protection program the licensee failed to adequately test the remote shutdown emergency transfer switch functions used to assure isolation of safe shutdown equipment from the control room in the event of a control room evacuation due to fire. The switch functions had not been adequately tested since 1997.
The failure to ensure isolation from the control room for safe shutdown equipment controlled from the remote shutdown panel during surveillance testing of emergency transfer switches is a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone in that it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown. Using Appendix F, , Degradation Rating Guidance Specific to Various Fire Protection Program Elements, the team determined that the finding constituted a low degradation of the safe shutdown area since the control room isolation feature was expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. This finding screened as having very low safety significance (Green). This violation was entered into the licensees corrective action program as CR-RBS-2010-01783. Because the emergency transfer switch surveillance procedures had been in effect since 1997, there was no crosscutting aspect associated with the violation, in that it is not indicative of current licensee performance. (Section 1R05.05.b.3)
Inspection Report# : 2010006 (pdf)
Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Evaluate a Revised Equipment Tag Out A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for defeating the Division I emergency systems automatic start functions caused by the failure to follow a work implementation and closeout procedure when changing the work scope and tag out boundaries for a safety-related maintenance activity. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-06151.
The finding was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to
 
prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the finding did not represent an actual loss of safety function. This finding has a crosscutting aspect in the area of human performance, work control because the licensee did not appropriately plan activities by incorporating actions to address operational impact and risk for the work scope changes [H.3(a)].
Inspection Report# : 2010002 (pdf)
Significance:      Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Control Building Chiller Operability During Low Service Water Temperatures The inspectors identified a Green noncited violation of Technical Specification 3.7.3 for exceeding the control room air conditioning system thirty day allowed outage time for one inoperable subsystem and the seven day allowed outage time for two inoperable subsystems and failing to enter Modes 3 and 4, as specified. Specifically, during accident conditions the control building chillers were not able to remove the design basis heat load while operating with low standby cooling water temperatures. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-01593, CR-RBS-2010-01817 and CR-RBS-2010-01667.
The performance deficiency was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issue Disposition Screening, because the finding was associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability of multiple safety-related systems and components to 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 because the finding resulted in an actual loss of safety function of a single train for greater than its technical specification allowed outage time and required a Phase 2 analysis. However, the Phase 2 presolved table and worksheets did not contain appropriate target sets to estimate accurately the risk impact of the finding. Therefore, the senior reactor analyst performed a Phase 3 analysis. The estimated change in core damage frequency was 2.3E-8/yr. Therefore, the inspectors determined the significance of the finding was Green. This finding was not assigned a crosscutting aspect because it does not reflect current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:      Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Adequately Verify a Suitable Replacement Part Essential for Emergency Diesel Generator Operation A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for the failure to adequately verify suitable replacement parts essential to the operation of emergency diesel generator Division I. This resulted in multiple intercooler flange bolts failing from low stress, high cycle fatigue. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-06148.
The finding is also similar to example 3j of Manual Chapter 0612 Appendix E. Specifically, the number of bolting failures placed the emergency diesel generators operability in doubt and an engineering analysis had to be performed to prove operability. In accordance with Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the inspectors performed a significant determination process Phase 1 screening and determined that the finding was of very low safety significance (Green) because a licensee analysis concluded that the bolts that were projected to fail during the emergency diesel generator mission time of thirty days would not result in an actual loss of system safety function. The inspectors determined that the finding had a crosscutting aspect in the area of human performance resources in that the licensee failed to ensure that equipment was adequate for maintaining long term plant safety by maintenance of design margins [H.2(a)].
Inspection Report# : 2010002 (pdf)
Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation
 
Failure to Control Scaffold Construction The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for the failure of maintenance personnel to control scaffold erection per procedure. This failure resulted in the licensee installing 31 scaffolds in safety related areas that required either rework or an engineering evaluation to resolve as built deviations from the minimum seismic separation requirements. As a result, the design function of the safety related equipment was potentially adversely affected. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-3963.
The failure to erect scaffolds in accordance with procedures is a performance deficiency. This finding is more than minor because it is similar to Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Section 4, Example a, because Entergy had routinely failed to perform the requisite engineering evaluation and because it was associated with the protection against external events 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 in order to prevent undesirable consequences. The finding was determined to be of very low risk significance (Green) because no actual loss of safety function occurred and the finding did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the area of human performance, work practices, because the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Reactor Core Isolation Cooling System Seismic Design The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to implement measures to ensure that the seismic design basis for the reactor core isolation cooling turbine governor hydraulic system was correctly translated into the specifications, drawings, procedures, or instructions. This resulted in work to reroute the piping and an engineering evaluation to resolve seismic concerns. The licensee entered this issue into their corrective action program as Condition Report CR RBS 2009 3747.
The failure to implement design control features for the seismic design of the reactor core isolation cooling system is a performance deficiency. This finding was more than minor because it is similar to Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, Section 5, Example a, in that the reactor core isolation cooling turbine was returned to service without the seismic spacing required by the original design or completion of an evaluation for the as left condition. This resulted in rework and additional engineering analysis to correctly resolve the seismic qualification concerns. The performance deficiency also affected the mitigating systems cornerstone attribute of external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Attachment A, Phase 1 - Initial Screening and Characterization of Findings, for the mitigating systems cornerstone. After answering no to all five questions in the mitigating systems cornerstone column of Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance. This finding does not have a crosscutting aspect because the performance deficiency occurred in 1989 and is not reflective of current plant performance.
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Standby Liquid Control System Test Tank Remained Drained The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for the failure of operations personnel to provide adequate procedural guidance to preclude water intrusion into the nonseismically qualified standby liquid control system test tank which resulted in the degradation of both trains of the standby liquid control system. The licensee entered this issue into their corrective action program as Condition Report CR RBS 2009 3862.
 
The failure to provide appropriate procedures to keep the standby liquid control test tank drained is a performance deficiency. The finding is more than minor because it affects the protection against external events attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance because the finding was not a design or qualification deficiency, did not represent a loss of a system/train safety function, and did not screen as potentially risk significant due to external events. This finding has a crosscutting aspect in the area of problem identification and resolutions corrective action program because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, the licensee failed to address the cause of inadvertent water intrusion into the standby liquid control test tank in a timely manner to prevent the common mode failure of both trains of standby liquid control [P.1(d)].
Inspection Report# : 2009004 (pdf)
Barrier Integrity Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Containment Closure Procedure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of operations personnel to fully implement a station procedure to control obstructions in primary containment openings in Modes 4 and 5. The failure to follow procedure challenged the licensees ability to establish containment closure. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-4296.
The failure to implement the containment closure procedure is a performance deficiency. This finding is more than minor because it affected the configuration control attribute of the barrier integrity objective to provide reasonable assurance that the physical design barriers (containment) will protect the public from radionuclide releases. Using Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, the finding was assessed as a Type B finding because it is related to a degraded condition that has potentially important implications for the integrity of the containment without affecting the likelihood of core damage and was of very low significance because the licensee did not lose the capability to close containment when planned. The finding has a crosscutting aspect in the area of human performance, work control, because the licensee failed to appropriately coordinate work activities (identifying cables, quick disconnects, removing unidentified cables) to address the operational impact of those work activities on containment operability [H.3(b)].
Inspection Report# : 2009004 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:        Dec 31, 2009 Identified By: NRC
 
Item Type: NCV NonCited Violation Failure to Ensure the Emergency Contact had Knowledge About a Shipment The inspectors identified a noncited violation of 10 CFR 71.5 and 49 CFR Part 172.604(a) for a failure to ensure that the shift manager, whose phone number was listed as the required 24-hour emergency phone number on shipping documents, was knowledgeable about the radioactive waste shipment that left site on December 16, 2009, and had immediate access to a person who had specific information on the shipment. Specifically, the shift manager was listed as the required 24-hour contact; however, the shift managers (on multiple shifts) were not provided with documentation or information about the shipments that left the site on December 16, 2009. Although the shift manager would have eventually contacted a knowledgeable person, this delay would not have resulted in immediate access to the person with information related to the shipment. The licensee immediately provided the shift manager a copy of the shipping documentation, briefed the shift manager, and entered this issue into their corrective action program as Condition Report CR-RBS-2009-06419.
This performance deficiency was more than minor because it adversely affected the public radiation safety cornerstone to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain in that the failure to have shipment information immediately available could restrict the actions of fire department and/or rescue personnel responding to an accident. When processed through the Public Radiation Safety Determination Process, the finding was determined to be of very low safety significance because the finding:
(1) was associated with radioactive material control, (2) involved the licensees program for radioactive material transportation, (3) did not cause radiation limits to be exceeded, (4) did not involve a breach of package during transit, (5) did not involve a certificate of compliance finding, (6) did not involve a low level burial ground nonconformance, and (7) did not involve a failure to make notifications. The inspectors determined the finding had a crosscutting aspect in area of resources, associated with documentation, because the licensees procedures did not provide guidance on informing the control room about shipments and thus, the procedures were not complete, accurate nor up-to-date [H.2 (c)].
Inspection Report# : 2009005 (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 : September 02, 2010
 
River Bend 1 3Q/2010 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Main Steam Line Plug Seal Failure Results in Loss of Reactor Cavity Inventory A self-revealing noncited violation of 10CFR50 Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the failure to follow the procedure for making a permanent plant modification and provide adequate procedures for installation and use of the main steam line plugs following a main steam line plug design change. This failure resulted in draining approximately 5,000 gallons of water from the upper reactor cavity pool to the drywell and a manual actuation of low pressure coolant injection to restore cavity pool water level. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-4681.
The finding was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issues Screening, because the finding affected the initiating events cornerstone attribute of configuration control and the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The inspectors evaluated the finding in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors reviewed Section II.B.(1) of Checklist 7 and determined that the finding required a Phase 2 analysis because the finding involved procedures that affected steam line plug seal configuration and resulted in inventory loss from the upper reactor cavity pool. The senior reactor analyst determined that, because of the special circumstances of this event, the use of a qualitative assessment using Inspection Manual Chapter 0609, Appendix M, was more appropriate than the risk tools provided in Inspection Manual Chapter 0609, Appendix G. This is because the draindown event was self-limiting, such that the inventory excursion could not have drained reactor cavity level below the level of the main steam lines, and that even with the failure of operators to take actions, the core would have remained covered with no challenges to the shutdown cooling system. Therefore, the event in the worst case would have been transparent to the core. Also, the displaced inventory posed no threat to any of the plant's mitigating systems. The inspectors concluded that the finding was of very low safety significance (Green). There is no crosscutting aspect associated with this violation because the finding does not reflect current licensee performance.
Inspection Report# : 2009005 (pdf)
Mitigating Systems Significance:        Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for High Pressure Core Spray Room Unit Cooler Maintenance The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the licensees failure to perform an adequate risk assessment while the high pressure core spray room unit cooler was unavailable. Specifically, the licensee assumed that risk would remain green and high pressure core spray would continue to inject into the reactor vessel for 6 hours after room cooling was made unavailable, when, in fact, risk became yellow because high pressure core spray would become unreliable after approximately 60 minutes due to instrument failure in the pumps minimum flow logic. As immediate corrective action, the licensee issued a standing order that administratively considered high pressure core spray unavailable when its room cooler is removed from service. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-02937.
 
This issue was more 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 Inspection Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time period is less than 1.0E-6 and because the licensee used incorrect risk assumptions that changed the outcome of their risk assessment. There is no crosscutting aspect associated with this violation because the assumptions that lead to the performance deficiency are not indicative of current licensee performance.
Inspection Report# : 2010003 (pdf)
Significance:      Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedure Results in Loss of Offsite Power to a 4160 Vac Safety Bus A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for inadequate procedural guidance when surveillance testing the Division III diesel generator. This resulted in a loss of offsite power to the Division III 4160 volt alternating current (Vac) bus while starting a nonsafety-related load with the Division III emergency diesel generator at full power paralleled to the grid. To prevent isolating offsite power to any safety bus, the licensee issued procedure changes to prevent starting loads on the nonsafety-related buses connected to the divisional safety buses while the emergency diesel generators are in test. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-00910.
The finding was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Table 4a, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its 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, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. There is no crosscutting aspect associated with this violation because this is a historical condition not previously identified by the licensee.
Inspection Report# : 2010003 (pdf)
Significance:      Jun 02, 2010 Identified By: NRC Item Type: VIO Violation Failure to Ensure at Least One Train of Equipment Necessary to Achieve Hot Shutdown Conditions is Free of Fire Damage The team identified a cited violation of License Condition 2.C.(10), Fire Protection, for failing to ensure that the Division 1 standby service water support system to the Division 1 emergency diesel generator, which was required to achieve safe shutdown, was protected such that it remained free from fire damage under all conditions. This condition was identified by the licensee in May 2007, and entered into their corrective action program as a significant non-conforming condition in CR-RBS-2007-02102. The licensee subsequently initiated compensatory measures in the form of manual actions to protect the Division 1 emergency diesel generator. This issue was documented as a licensee-identified noncited violation in Inspection Report 2009002. River Bend has subsequently completed two refueling outages, six forced outages, and one emergency diesel generator work window of sufficient duration since identification of this condition and failed to correct the non-conformance. The team determined that schedule changes resulted in a new completion date of January 2011.
The failure to ensure that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) was free of fire damage and to correct this significant non-conforming condition in a timely manner is a performance deficiency. This performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems
 
Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A Phase 3 SDP risk assessment was performed by a senior reactor analyst. The bounding change in conditional core damage frequency for a 1-year exposure is the Fire Mitigation Frequency (4.30E-08/year) multiplied by the change in conditional core damage probability (0.9) for a value of 3.87E-08/year. This value indicates the finding has very low safety significance (Green). Because the licensee failed to correct this violation, this violation is being treated as a cited violation, consistent with the NRC Enforcement Policy. This finding had a crosscutting aspect in the Work Control component of the Human Performance area because the licensee did not appropriately plan work activities to support long-term equipment reliability by limiting temporary modifications, operator workarounds, safety systems unavailability, and reliance on manual actions [H.3(b)]. (Section 1R05.01)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Alternative Shutdown Procedure Could be Implemented as Written The team identified a noncited violation of Technical Specification 5.4.1.d, Fire Protection Program Implementation. Specifically, Procedure AOP-0031 Shutdown from Outside the Main Control Room, Revision 307, had steps that could not be implemented as written. Two steps were to be performed before the necessary ac power was available, and two steps required diagnostic assessment without the availability of instrumentation.
The failure to ensure that Procedure AOP-0031, Revision 307 could be implemented as written is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Attachment 2 to Appendix F, Fire Protection Significance Determination Process, this issue was determined to be a safe shutdown finding, and was assigned a degradation rating of Low because the examples involved procedural deficiencies that could be compensated for by operator experience. Since this finding was assigned a low degradation rating, the safety significance screened as very low (Green). This finding was entered into the licensees corrective action program as CR-RBS-2010-01592, CR-RBS-2010-01831, CR-RBS-2010-01775, CR-RBS-2010-01821, and CR-RBS-2010-1846. This finding had a crosscutting aspect in the Resources component of the Human Performance area, in that the licensee did not ensure that procedures were complete, accurate, and up to date to assure nuclear safety [H.2.(c)]. (Section 1R05.05.b.1)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement and Maintain in Effect all Provisions of the Approved Fire Protection Program The team identified a noncited violation of License Condition 2.C.(10), Fire Protection, for the failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the team identified, during a timed walkdown of the procedure that it took operators over 6 minutes to isolate feedwater, but the simulator showed that the steam lines could be flooded in 2 minutes. Overfilling the reactor pressure vessel and flooding the main steam lines could make reactor core isolation cooling unavailable. Reactor core isolation cooling was credited for decay heat removal and inventory control in the event of a fire.
The failure to ensure that feedwater would be isolated prior to overfilling the reactor pressure vessel and flooding the main steam lines making reactor core isolation cooling unavailable is a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team
 
evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a control room fire that led to control room abandonment. The Phase 3 evaluation determined that the finding had very low safety significance because a fire in only one of 109 electrical cabinets in the control room could result in this overfill event. The finding was entered into the licensees corrective action program as CR-RBS-2010-01808. The finding did not have a crosscutting aspect since it was not indicative of current performance, in that the licensee had established the incorrect response time more than three years prior to this finding. (Section 1R05.05.b.2)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement and Maintain in Effect all Provisions of the Approved Fire Protection Program The team identified a noncited violation of License Condition 2.C.(10), Fire Protection, related to the licensee's failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, during testing required by the approved fire protection program the licensee failed to adequately test the remote shutdown emergency transfer switch functions used to assure isolation of safe shutdown equipment from the control room in the event of a control room evacuation due to fire. The switch functions had not been adequately tested since 1997.
The failure to ensure isolation from the control room for safe shutdown equipment controlled from the remote shutdown panel during surveillance testing of emergency transfer switches is a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone in that it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown. Using Appendix F, , Degradation Rating Guidance Specific to Various Fire Protection Program Elements, the team determined that the finding constituted a low degradation of the safe shutdown area since the control room isolation feature was expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. This finding screened as having very low safety significance (Green). This violation was entered into the licensees corrective action program as CR-RBS-2010-01783. Because the emergency transfer switch surveillance procedures had been in effect since 1997, there was no crosscutting aspect associated with the violation, in that it is not indicative of current licensee performance. (Section 1R05.05.b.3)
Inspection Report# : 2010006 (pdf)
Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Evaluate a Revised Equipment Tag Out A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for defeating the Division I emergency systems automatic start functions caused by the failure to follow a work implementation and closeout procedure when changing the work scope and tag out boundaries for a safety-related maintenance activity. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-06151.
The finding was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the finding did not represent an actual loss of safety function. This finding has a crosscutting aspect in the area of human performance, work control because the licensee did not appropriately plan activities by incorporating actions to address operational impact and risk for the work scope changes [H.3(a)].
Inspection Report# : 2010002 (pdf)
 
Significance:      Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Control Building Chiller Operability During Low Service Water Temperatures The inspectors identified a Green noncited violation of Technical Specification 3.7.3 for exceeding the control room air conditioning system thirty day allowed outage time for one inoperable subsystem and the seven day allowed outage time for two inoperable subsystems and failing to enter Modes 3 and 4, as specified. Specifically, during accident conditions the control building chillers were not able to remove the design basis heat load while operating with low standby cooling water temperatures. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-01593, CR-RBS-2010-01817 and CR-RBS-2010-01667.
The performance deficiency was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issue Disposition Screening, because the finding was associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability of multiple safety-related systems and components to 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 because the finding resulted in an actual loss of safety function of a single train for greater than its technical specification allowed outage time and required a Phase 2 analysis. However, the Phase 2 presolved table and worksheets did not contain appropriate target sets to estimate accurately the risk impact of the finding. Therefore, the senior reactor analyst performed a Phase 3 analysis. The estimated change in core damage frequency was 2.3E-8/yr. Therefore, the inspectors determined the significance of the finding was Green. This finding was not assigned a crosscutting aspect because it does not reflect current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:      Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Adequately Verify a Suitable Replacement Part Essential for Emergency Diesel Generator Operation A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for the failure to adequately verify suitable replacement parts essential to the operation of emergency diesel generator Division I. This resulted in multiple intercooler flange bolts failing from low stress, high cycle fatigue. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-06148.
The finding is also similar to example 3j of Manual Chapter 0612 Appendix E. Specifically, the number of bolting failures placed the emergency diesel generators operability in doubt and an engineering analysis had to be performed to prove operability. In accordance with Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the inspectors performed a significant determination process Phase 1 screening and determined that the finding was of very low safety significance (Green) because a licensee analysis concluded that the bolts that were projected to fail during the emergency diesel generator mission time of thirty days would not result in an actual loss of system safety function. The inspectors determined that the finding had a crosscutting aspect in the area of human performance resources in that the licensee failed to ensure that equipment was adequate for maintaining long term plant safety by maintenance of design margins [H.2(a)].
Inspection Report# : 2010002 (pdf)
Barrier Integrity Emergency Preparedness
 
Occupational Radiation Safety Significance:      Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit Instructions NRC inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1 for failure to follow radiation work permit instructions. Specifically, a team technician made an unauthorized entry into a posted high radiation area on a radiation work permit that did not grant access to that area. The licensee conducted a review of this event and issued a site-wide memorandum on procedural and management expectations associated with high radiation areas. The licensee entered the finding into the corrective action program as Condition Report CR RBS-2009-03953.
The failure to follow the instructions on a radiation work permit is a performance deficiency. The performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone. It is associated with the exposure control attribute in that a worker not following radiation work permit instructions does not ensure adequate protection of the workers health and safety from additional/unintended personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Furthermore, the finding had an associated human performance crosscutting aspect in the work practices component because the worker did not use human error prevention techniques, such as self-checking [H.4(a)].
Inspection Report# : 2010003 (pdf)
Public Radiation Safety Significance:      Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure the Emergency Contact had Knowledge About a Shipment The inspectors identified a noncited violation of 10 CFR 71.5 and 49 CFR Part 172.604(a) for a failure to ensure that the shift manager, whose phone number was listed as the required 24-hour emergency phone number on shipping documents, was knowledgeable about the radioactive waste shipment that left site on December 16, 2009, and had immediate access to a person who had specific information on the shipment. Specifically, the shift manager was listed as the required 24-hour contact; however, the shift managers (on multiple shifts) were not provided with documentation or information about the shipments that left the site on December 16, 2009. Although the shift manager would have eventually contacted a knowledgeable person, this delay would not have resulted in immediate access to the person with information related to the shipment. The licensee immediately provided the shift manager a copy of the shipping documentation, briefed the shift manager, and entered this issue into their corrective action program as Condition Report CR-RBS-2009-06419.
This performance deficiency was more than minor because it adversely affected the public radiation safety cornerstone to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain in that the failure to have shipment information immediately available could restrict the actions of fire department and/or rescue personnel responding to an accident. When processed through the Public Radiation Safety Determination Process, the finding was determined to be of very low safety significance because the finding:
(1) was associated with radioactive material control, (2) involved the licensees program for radioactive material transportation, (3) did not cause radiation limits to be exceeded, (4) did not involve a breach of package during transit, (5) did not involve a certificate of compliance finding, (6) did not involve a low level burial ground nonconformance, and (7) did not involve a failure to make notifications. The inspectors determined the finding had a crosscutting aspect in area of resources, associated with documentation, because the licensees procedures did not provide guidance on informing the control room about shipments and thus, the procedures were not complete, accurate nor up-to-date [H.2 (c)].
 
Inspection Report# : 2009005 (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 : November 29, 2010
 
River Bend 1 4Q/2010 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding Failure to Plug a Main Condenser Tube in Accordance with an Approved Work Order The inspectors reviewed a self-revealing finding for the licensees failure to plug a main condenser tube in accordance with an approved work order. Specifically, a plastic tube plug was not replaced with the required brass plug causing a tube leak requiring the plant to reduce power. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-04526.
The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, in that the performance deficiency created a condition that upset plant stability by creating a condenser tube leak that prompted the plant to reduce power. The inspectors determined that the apparent cause of this finding was the licensees failure to use human performance error-prevention techniques to ensure that the tube plugging was performed correctly. This finding therefore has a crosscutting aspect in the work practices component of the human performance area because the licensee did not communicate and use human error prevention techniques commensurate with the risk of the assigned task, such that work activities are performed safely [(H.4(a)].
Inspection Report# : 2010005 (pdf)
Significance:      Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Licensed Operator Examination Integrity The inspectors identified a noncited violation of 10 CFR Part 55.49, Integrity of Examinations and Tests, for the failure of operations training personnel to ensure the integrity of an operating test administered to a licensed operations crew was maintained. One licensed operations crew received two scenarios for their operating test that had been previously administered to a licensed operations staff crew. This failure resulted in a compromise of examination integrity, but did not lead to an actual effect on the equitable and consistent administration of the examination. This finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions when adopting a 50 percent operating examination overlap practice
[H.1(b)].
The finding is more than minor because, if left uncorrected, the finding could have become more significant in that allowing untested licensed operators at the controls could be a precursor to a significant event if undetected performance deficiencies develop. The finding was determined to have very low safety significance (Green) because the finding resulted in a compromise of the integrity of operating test scenarios and compensatory actions were not immediately taken when the compromise should have been discovered. However, the equitable and consistent administration of the exam was not actually impacted by this compromise. The inspectors applied Inspection Manual Chapter 0609, Significance Determination Process, Appendix I, Licensed Operator Requalification Significance Determination Process, and determined that the finding should be dispositioned as a Green noncited violation.
Inspection Report# : 2010004 (pdf)
Mitigating Systems
 
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Develop a Preventive Maintenance Schedule to Specify Inspection or Replacement of the O-Ring in the High Pressure Core Spray Lower Motor Bearing Drain Plug The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1 for the licensees failure to determine the appropriate preventive maintenance strategy and task frequency for the o-ring in the high pressure core spray lower motor bearing drain plug. As immediate correction action, the licensee replaced the o-ring. At the conclusion of the inspection, the licensee was in the process of determining the appropriate replacement frequency.
The licensee entered this issue into their corrective action system as Condition Report CR-RBS-2010-05766.
This finding was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern, in that if the licensee did not develop a preventive maintenance schedule for periodically replacing the subject o-ring, degradation of that o-ring due to aging could allow a leak that would drain oil from the lower motor bearing and thus render the high pressure core spray pump inoperable. As described in Inspection Manual 0609 Appendix A, a Phase 2 analysis using the presolved worksheet determined that this finding had very low (Green) risk significance. This finding has a crosscutting aspect in the operating experience component of the problem identification & resolution area because the licensee did not systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience [P.2(a)].
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Two Examples of Completing Maintenance that Affected the Performance of Safety-Related Equipment but Was Not Properly Preplanned The inspectors reviewed a two-example self-revealing green noncited violation of Technical Specification 5.4.1 for two occasions on which the licensee completed maintenance that affected the performance of safety-related equipment (high pressure core spray) but was not properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances As a result, the licensee overtorqued the high pressure core spray lower motor bearing drain plug causing the plug to fracture. This fracture resulted in excessive oil leakage that caused the pump to become inoperable. The violation is in the licensees corrective action program as Condition Report CR-RBS-2011-00224.
These performance deficiencies were more than minor and therefore constituted a finding because they affected the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. As described in Inspection Manual 0609 Appendix A, a Phase 2 analysis using the presolved worksheet determined that this finding had very low risk significance. The finding has a crosscutting aspect in the resources component of the human performance area because the apparent cause of the finding was a procedure that was not adequate to assure nuclear safety [H.2(c)].
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Required Quality Control Inspections Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. The licensee performed extensive reviews, and inspectors performed independent reviews of the licensees conclusions as well as independent sampling, to confirm that improper or missed inspections did not actually affect the operability of plant equipment. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective
 
action program under Condition Reports CR-HQN-2009-01184 and CR-HQN-2010-0013.
The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the design control attribute of the Mitigating Systems Cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to have very low safety significance in Phase 1 of the significance determination process since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this performance deficiency involved a crosscutting aspect related to the human performance area associated with decision making [H.1 (a)] because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate.
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program.
Specifically, the individual assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as Condition Report CR-HQN-2010-00386.
Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems Cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no crosscutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago.
Inspection Report# : 2010005 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate High Pressure Core Spray Pump Room Cooler Bearing Maintenance A self-revealing, very low safety significance (Green) noncited violation 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was reviewed for the licensees failure to prescribe lubrication and installation of bearings on the high-pressure core spray room cooler motors by adequate procedures. In response to this finding, the licensee changed their procedure for performing material equivalency evaluations to require that, when plant components change and associated vendor-recommended maintenance schedules change, licensee personnel also update the corresponding preventive-maintenance tasks. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-02919.
This finding was more 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, in that this finding caused inoperability of the high-pressure core spray. The significance of this finding was determined by completing a Phase 3 analysis in accordance with Inspection Manual Chapter 0609, Appendix A, which determined that the incremental core damage probability maximum was 2x10-7, and that the finding was therefore of very low safety significance (Green). This finding did not represent current licensee performance and consequently did not have a cross-cutting aspect because the cause of this finding was that when the licensee replaced a component by a similar component from a different vendor, no licensee procedure required them to update the associated maintenance frequencies, and because before this finding was identified, the licensee had no reasonable opportunity to identify and correct that deficiency in that procedure.
Inspection Report# : 2010004 (pdf)
Significance:      Sep 30, 2010 Identified By: NRC Item Type: FIN Finding Inadequate Maintenance Results in Unplanned Opening of Main Turbine Bypass Valve A self-revealing finding of very low safety significance (Green) was identified when turbine bypass valve number 1 opened unexpectedly causing the reactor to exceed 100 percent core thermal power. Operators promptly lowered core thermal power to 90 percent to preserve margin to fuel thermal limits. A failed power supply and inadequate calibration and testing of the steam bypass and pressure regulation system and electro-hydraulic control system caused the event. Corrective actions include replacing system power supplies and revising applicable calibration and test instructions. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-03343.
This finding was more 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. Specifically, technical specification mitigation equipment (main turbine bypass system, end-of-cycle recirculation pump trip function, and rod block instrumentation functions) became inoperable. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that the finding was of very low safety significance (Green) because the finding 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 this finding did not represent current licensee performance because the preventative maintenance schedule and calibration procedure were developed and approved over two years ago. Therefore, no crosscutting aspect was assigned to this finding.
Inspection Report# : 2010004 (pdf)
Significance:      Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for High Pressure Core Spray Room Unit Cooler Maintenance The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the licensees failure to perform an adequate risk assessment while the high pressure core spray room unit cooler was unavailable. Specifically, the licensee assumed that risk would remain green and high pressure core spray would continue to inject into the reactor vessel for 6 hours after room cooling was made unavailable, when, in fact, risk became yellow because high pressure core spray would become unreliable after approximately 60 minutes due to instrument failure in the pumps minimum flow logic. As immediate corrective action, the licensee issued a standing order that administratively considered high pressure core spray unavailable when its room cooler is removed from service. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-02937.
This issue was more 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 Inspection Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time period is less than 1.0E-6 and
 
because the licensee used incorrect risk assumptions that changed the outcome of their risk assessment. There is no crosscutting aspect associated with this violation because the assumptions that lead to the performance deficiency are not indicative of current licensee performance.
Inspection Report# : 2010003 (pdf)
Significance:      Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedure Results in Loss of Offsite Power to a 4160 Vac Safety Bus A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for inadequate procedural guidance when surveillance testing the Division III diesel generator. This resulted in a loss of offsite power to the Division III 4160 volt alternating current (Vac) bus while starting a nonsafety-related load with the Division III emergency diesel generator at full power paralleled to the grid. To prevent isolating offsite power to any safety bus, the licensee issued procedure changes to prevent starting loads on the nonsafety-related buses connected to the divisional safety buses while the emergency diesel generators are in test. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-00910.
The finding was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Table 4a, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its 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, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. There is no crosscutting aspect associated with this violation because this is a historical condition not previously identified by the licensee.
Inspection Report# : 2010003 (pdf)
Significance:      Jun 02, 2010 Identified By: NRC Item Type: VIO Violation Failure to Ensure at Least One Train of Equipment Necessary to Achieve Hot Shutdown Conditions is Free of Fire Damage The team identified a cited violation of License Condition 2.C.(10), Fire Protection, for failing to ensure that the Division 1 standby service water support system to the Division 1 emergency diesel generator, which was required to achieve safe shutdown, was protected such that it remained free from fire damage under all conditions. This condition was identified by the licensee in May 2007, and entered into their corrective action program as a significant non-conforming condition in CR-RBS-2007-02102. The licensee subsequently initiated compensatory measures in the form of manual actions to protect the Division 1 emergency diesel generator. This issue was documented as a licensee-identified noncited violation in Inspection Report 2009002. River Bend has subsequently completed two refueling outages, six forced outages, and one emergency diesel generator work window of sufficient duration since identification of this condition and failed to correct the non-conformance. The team determined that schedule changes resulted in a new completion date of January 2011.
The failure to ensure that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) was free of fire damage and to correct this significant non-conforming condition in a timely manner is a performance deficiency. This performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A Phase 3 SDP risk assessment was performed by a senior reactor analyst. The bounding change in conditional core damage frequency for a 1-year exposure is the Fire Mitigation Frequency (4.30E-08/year)
 
multiplied by the change in conditional core damage probability (0.9) for a value of 3.87E-08/year. This value indicates the finding has very low safety significance (Green). Because the licensee failed to correct this violation, this violation is being treated as a cited violation, consistent with the NRC Enforcement Policy. This finding had a crosscutting aspect in the Work Control component of the Human Performance area because the licensee did not appropriately plan work activities to support long-term equipment reliability by limiting temporary modifications, operator workarounds, safety systems unavailability, and reliance on manual actions [H.3(b)]. (Section 1R05.01)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Alternative Shutdown Procedure Could be Implemented as Written The team identified a noncited violation of Technical Specification 5.4.1.d, Fire Protection Program Implementation. Specifically, Procedure AOP-0031 Shutdown from Outside the Main Control Room, Revision 307, had steps that could not be implemented as written. Two steps were to be performed before the necessary ac power was available, and two steps required diagnostic assessment without the availability of instrumentation.
The failure to ensure that Procedure AOP-0031, Revision 307 could be implemented as written is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Attachment 2 to Appendix F, Fire Protection Significance Determination Process, this issue was determined to be a safe shutdown finding, and was assigned a degradation rating of Low because the examples involved procedural deficiencies that could be compensated for by operator experience. Since this finding was assigned a low degradation rating, the safety significance screened as very low (Green). This finding was entered into the licensees corrective action program as CR-RBS-2010-01592, CR-RBS-2010-01831, CR-RBS-2010-01775, CR-RBS-2010-01821, and CR-RBS-2010-1846. This finding had a crosscutting aspect in the Resources component of the Human Performance area, in that the licensee did not ensure that procedures were complete, accurate, and up to date to assure nuclear safety [H.2.(c)]. (Section 1R05.05.b.1)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement and Maintain in Effect all Provisions of the Approved Fire Protection Program The team identified a noncited violation of License Condition 2.C.(10), Fire Protection, for the failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the team identified, during a timed walkdown of the procedure that it took operators over 6 minutes to isolate feedwater, but the simulator showed that the steam lines could be flooded in 2 minutes. Overfilling the reactor pressure vessel and flooding the main steam lines could make reactor core isolation cooling unavailable. Reactor core isolation cooling was credited for decay heat removal and inventory control in the event of a fire.
The failure to ensure that feedwater would be isolated prior to overfilling the reactor pressure vessel and flooding the main steam lines making reactor core isolation cooling unavailable is a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a control room fire that led to control room abandonment. The Phase 3 evaluation determined that the finding had very low safety significance because a fire in only one of 109 electrical cabinets in the control room could result in this overfill event. The finding was entered into
 
the licensees corrective action program as CR-RBS-2010-01808. The finding did not have a crosscutting aspect since it was not indicative of current performance, in that the licensee had established the incorrect response time more than three years prior to this finding. (Section 1R05.05.b.2)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement and Maintain in Effect all Provisions of the Approved Fire Protection Program The team identified a noncited violation of License Condition 2.C.(10), Fire Protection, related to the licensee's failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, during testing required by the approved fire protection program the licensee failed to adequately test the remote shutdown emergency transfer switch functions used to assure isolation of safe shutdown equipment from the control room in the event of a control room evacuation due to fire. The switch functions had not been adequately tested since 1997.
The failure to ensure isolation from the control room for safe shutdown equipment controlled from the remote shutdown panel during surveillance testing of emergency transfer switches is a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone in that it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown. Using Appendix F, , Degradation Rating Guidance Specific to Various Fire Protection Program Elements, the team determined that the finding constituted a low degradation of the safe shutdown area since the control room isolation feature was expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. This finding screened as having very low safety significance (Green). This violation was entered into the licensees corrective action program as CR-RBS-2010-01783. Because the emergency transfer switch surveillance procedures had been in effect since 1997, there was no crosscutting aspect associated with the violation, in that it is not indicative of current licensee performance. (Section 1R05.05.b.3)
Inspection Report# : 2010006 (pdf)
Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Evaluate a Revised Equipment Tag Out A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for defeating the Division I emergency systems automatic start functions caused by the failure to follow a work implementation and closeout procedure when changing the work scope and tag out boundaries for a safety-related maintenance activity. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-06151.
The finding was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the finding did not represent an actual loss of safety function. This finding has a crosscutting aspect in the area of human performance, work control because the licensee did not appropriately plan activities by incorporating actions to address operational impact and risk for the work scope changes [H.3(a)].
Inspection Report# : 2010002 (pdf)
Significance:        Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Control Building Chiller Operability During Low Service Water Temperatures The inspectors identified a Green noncited violation of Technical Specification 3.7.3 for exceeding the control room air conditioning system thirty day allowed outage time for one inoperable subsystem and the seven day allowed
 
outage time for two inoperable subsystems and failing to enter Modes 3 and 4, as specified. Specifically, during accident conditions the control building chillers were not able to remove the design basis heat load while operating with low standby cooling water temperatures. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-01593, CR-RBS-2010-01817 and CR-RBS-2010-01667.
The performance deficiency was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issue Disposition Screening, because the finding was associated with the mitigating systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability of multiple safety-related systems and components to 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 because the finding resulted in an actual loss of safety function of a single train for greater than its technical specification allowed outage time and required a Phase 2 analysis. However, the Phase 2 presolved table and worksheets did not contain appropriate target sets to estimate accurately the risk impact of the finding. Therefore, the senior reactor analyst performed a Phase 3 analysis. The estimated change in core damage frequency was 2.3E-8/yr. Therefore, the inspectors determined the significance of the finding was Green. This finding was not assigned a crosscutting aspect because it does not reflect current licensee performance.
Inspection Report# : 2010002 (pdf)
Significance:      Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Adequately Verify a Suitable Replacement Part Essential for Emergency Diesel Generator Operation A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for the failure to adequately verify suitable replacement parts essential to the operation of emergency diesel generator Division I. This resulted in multiple intercooler flange bolts failing from low stress, high cycle fatigue. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2009-06148.
The finding is also similar to example 3j of Manual Chapter 0612 Appendix E. Specifically, the number of bolting failures placed the emergency diesel generators operability in doubt and an engineering analysis had to be performed to prove operability. In accordance with Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the inspectors performed a significant determination process Phase 1 screening and determined that the finding was of very low safety significance (Green) because a licensee analysis concluded that the bolts that were projected to fail during the emergency diesel generator mission time of thirty days would not result in an actual loss of system safety function. The inspectors determined that the finding had a crosscutting aspect in the area of human performance resources in that the licensee failed to ensure that equipment was adequate for maintaining long term plant safety by maintenance of design margins [H.2(a)].
Inspection Report# : 2010002 (pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance:      Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation
 
Failure to Follow Radiation Work Permit Instructions NRC inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1 for failure to follow radiation work permit instructions. Specifically, a team technician made an unauthorized entry into a posted high radiation area on a radiation work permit that did not grant access to that area. The licensee conducted a review of this event and issued a site-wide memorandum on procedural and management expectations associated with high radiation areas. The licensee entered the finding into the corrective action program as Condition Report CR RBS-2009-03953.
The failure to follow the instructions on a radiation work permit is a performance deficiency. The performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone. It is associated with the exposure control attribute in that a worker not following radiation work permit instructions does not ensure adequate protection of the workers health and safety from additional/unintended personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Furthermore, the finding had an associated human performance crosscutting aspect in the work practices component because the worker did not use human error prevention techniques, such as self-checking [H.4(a)].
Inspection Report# : 2010003 (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 Significance: TBD Sep 30, 2010 Identified By: NRC Item Type: AV Apparent Violation Failure to provide adequate decommissioning funding assurance from Dec. 31, 2008 through the present (10 CFR 50.75(b))
To determine the licensees compliance with the decommissioning financial assurance requirements in 10 CFR 50.75, the Nuclear Regulatory Commission reviewed the decommissioning funding report for the River Bend Station, submitted by the licensee on March 30, 2009. The review is described in ADAMS document ML1025200621 and identified three apparent violations.
Inspection Report# : 2010004 (pdf)
Significance: TBD Sep 30, 2010 Identified By: NRC Item Type: AV Apparent Violation Failure to provide complete and accurate information by failing to disclose its reliance on a contract in the March 2009 funds status report (10 CFR 50.75(f)(1) and 10 CFR 50.9)
To determine the licensees compliance with the decommissioning financial assurance requirements in 10 CFR 50.75, the Nuclear Regulatory Commission reviewed the decommissioning funding report for the River Bend Station, submitted by the licensee on March 30, 2009. The review is described in ADAMS document ML1025200621 and identified three apparent violations.
Inspection Report# : 2010004 (pdf)
 
Significance: TBD Sep 30, 2010 Identified By: NRC Item Type: AV Apparent Violation Use of a decommissioning funding mechanism that did not meet the requirements of 10 CFR 50.75(3)(1)(v)
To determine the licensees compliance with the decommissioning financial assurance requirements in 10 CFR 50.75, the Nuclear Regulatory Commission reviewed the decommissioning funding report for the River Bend Station, submitted by the licensee on March 30, 2009. The review is described in ADAMS document ML1025200621 and identified three apparent violations.
Inspection Report# : 2010004 (pdf)
Last modified : March 03, 2011
 
River Bend 1 1Q/2011 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding Failure to Plug a Main Condenser Tube in Accordance with an Approved Work Order The inspectors reviewed a self-revealing finding for the licensees failure to plug a main condenser tube in accordance with an approved work order. Specifically, a plastic tube plug was not replaced with the required brass plug causing a tube leak requiring the plant to reduce power. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-04526.
The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, in that the performance deficiency created a condition that upset plant stability by creating a condenser tube leak that prompted the plant to reduce power. The inspectors determined that the apparent cause of this finding was the licensees failure to use human performance error-prevention techniques to ensure that the tube plugging was performed correctly. This finding therefore has a crosscutting aspect in the work practices component of the human performance area because the licensee did not communicate and use human error prevention techniques commensurate with the risk of the assigned task, such that work activities are performed safely [(H.4(a)].
Inspection Report# : 2010005 (pdf)
Significance:      Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Licensed Operator Examination Integrity The inspectors identified a noncited violation of 10 CFR Part 55.49, Integrity of Examinations and Tests, for the failure of operations training personnel to ensure the integrity of an operating test administered to a licensed operations crew was maintained. One licensed operations crew received two scenarios for their operating test that had been previously administered to a licensed operations staff crew. This failure resulted in a compromise of examination integrity, but did not lead to an actual effect on the equitable and consistent administration of the examination. This finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions when adopting a 50 percent operating examination overlap practice
[H.1(b)].
The finding is more than minor because, if left uncorrected, the finding could have become more significant in that allowing untested licensed operators at the controls could be a precursor to a significant event if undetected performance deficiencies develop. The finding was determined to have very low safety significance (Green) because the finding resulted in a compromise of the integrity of operating test scenarios and compensatory actions were not immediately taken when the compromise should have been discovered. However, the equitable and consistent administration of the exam was not actually impacted by this compromise. The inspectors applied Inspection Manual Chapter 0609, Significance Determination Process, Appendix I, Licensed Operator Requalification Significance Determination Process, and determined that the finding should be dispositioned as a Green noncited violation.
Inspection Report# : 2010004 (pdf)
Mitigating Systems
 
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Develop a Preventive Maintenance Schedule to Specify Inspection or Replacement of the O-Ring in the High Pressure Core Spray Lower Motor Bearing Drain Plug The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1 for the licensees failure to determine the appropriate preventive maintenance strategy and task frequency for the o-ring in the high pressure core spray lower motor bearing drain plug. As immediate correction action, the licensee replaced the o-ring. At the conclusion of the inspection, the licensee was in the process of determining the appropriate replacement frequency.
The licensee entered this issue into their corrective action system as Condition Report CR-RBS-2010-05766.
This finding was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern, in that if the licensee did not develop a preventive maintenance schedule for periodically replacing the subject o-ring, degradation of that o-ring due to aging could allow a leak that would drain oil from the lower motor bearing and thus render the high pressure core spray pump inoperable. As described in Inspection Manual 0609 Appendix A, a Phase 2 analysis using the presolved worksheet determined that this finding had very low (Green) risk significance. This finding has a crosscutting aspect in the operating experience component of the problem identification & resolution area because the licensee did not systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience [P.2(a)].
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Two Examples of Completing Maintenance that Affected the Performance of Safety-Related Equipment but Was Not Properly Preplanned The inspectors reviewed a two-example self-revealing green noncited violation of Technical Specification 5.4.1 for two occasions on which the licensee completed maintenance that affected the performance of safety-related equipment (high pressure core spray) but was not properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances As a result, the licensee overtorqued the high pressure core spray lower motor bearing drain plug causing the plug to fracture. This fracture resulted in excessive oil leakage that caused the pump to become inoperable. The violation is in the licensees corrective action program as Condition Report CR-RBS-2011-00224.
These performance deficiencies were more than minor and therefore constituted a finding because they affected the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. As described in Inspection Manual 0609 Appendix A, a Phase 2 analysis using the presolved worksheet determined that this finding had very low risk significance. The finding has a crosscutting aspect in the resources component of the human performance area because the apparent cause of the finding was a procedure that was not adequate to assure nuclear safety [H.2(c)].
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Required Quality Control Inspections Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. The licensee performed extensive reviews, and inspectors performed independent reviews of the licensees conclusions as well as independent sampling, to confirm that improper or missed inspections did not actually affect the operability of plant equipment. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective
 
action program under Condition Reports CR-HQN-2009-01184 and CR-HQN-2010-0013.
The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the design control attribute of the Mitigating Systems Cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to have very low safety significance in Phase 1 of the significance determination process since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this performance deficiency involved a crosscutting aspect related to the human performance area associated with decision making [H.1 (a)] because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate.
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program.
Specifically, the individual assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as Condition Report CR-HQN-2010-00386.
Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems Cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no crosscutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago.
Inspection Report# : 2010005 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate High Pressure Core Spray Pump Room Cooler Bearing Maintenance A self-revealing, very low safety significance (Green) noncited violation 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was reviewed for the licensees failure to prescribe lubrication and installation of bearings on the high-pressure core spray room cooler motors by adequate procedures. In response to this finding, the licensee changed their procedure for performing material equivalency evaluations to require that, when plant components change and associated vendor-recommended maintenance schedules change, licensee personnel also update the corresponding preventive-maintenance tasks. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-02919.
This finding was more 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, in that this finding caused inoperability of the high-pressure core spray. The significance of this finding was determined by completing a Phase 3 analysis in accordance with Inspection Manual Chapter 0609, Appendix A, which determined that the incremental core damage probability maximum was 2x10-7, and that the finding was therefore of very low safety significance (Green). This finding did not represent current licensee performance and consequently did not have a cross-cutting aspect because the cause of this finding was that when the licensee replaced a component by a similar component from a different vendor, no licensee procedure required them to update the associated maintenance frequencies, and because before this finding was identified, the licensee had no reasonable opportunity to identify and correct that deficiency in that procedure.
Inspection Report# : 2010004 (pdf)
Significance:      Sep 30, 2010 Identified By: NRC Item Type: FIN Finding Inadequate Maintenance Results in Unplanned Opening of Main Turbine Bypass Valve A self-revealing finding of very low safety significance (Green) was identified when turbine bypass valve number 1 opened unexpectedly causing the reactor to exceed 100 percent core thermal power. Operators promptly lowered core thermal power to 90 percent to preserve margin to fuel thermal limits. A failed power supply and inadequate calibration and testing of the steam bypass and pressure regulation system and electro-hydraulic control system caused the event. Corrective actions include replacing system power supplies and revising applicable calibration and test instructions. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-03343.
This finding was more 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. Specifically, technical specification mitigation equipment (main turbine bypass system, end-of-cycle recirculation pump trip function, and rod block instrumentation functions) became inoperable. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that the finding was of very low safety significance (Green) because the finding 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 this finding did not represent current licensee performance because the preventative maintenance schedule and calibration procedure were developed and approved over two years ago. Therefore, no crosscutting aspect was assigned to this finding.
Inspection Report# : 2010004 (pdf)
Significance:      Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for High Pressure Core Spray Room Unit Cooler Maintenance The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the licensees failure to perform an adequate risk assessment while the high pressure core spray room unit cooler was unavailable. Specifically, the licensee assumed that risk would remain green and high pressure core spray would continue to inject into the reactor vessel for 6 hours after room cooling was made unavailable, when, in fact, risk became yellow because high pressure core spray would become unreliable after approximately 60 minutes due to instrument failure in the pumps minimum flow logic. As immediate corrective action, the licensee issued a standing order that administratively considered high pressure core spray unavailable when its room cooler is removed from service. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-02937.
This issue was more 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 Inspection Manual Chapter 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is determined to have very low safety significance (Green) because the incremental core damage probability deficit for the affected time period is less than 1.0E-6 and
 
because the licensee used incorrect risk assumptions that changed the outcome of their risk assessment. There is no crosscutting aspect associated with this violation because the assumptions that lead to the performance deficiency are not indicative of current licensee performance.
Inspection Report# : 2010003 (pdf)
Significance:      Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedure Results in Loss of Offsite Power to a 4160 Vac Safety Bus A self-revealing noncited violation of Technical Specification 5.4.1.a. was identified for inadequate procedural guidance when surveillance testing the Division III diesel generator. This resulted in a loss of offsite power to the Division III 4160 volt alternating current (Vac) bus while starting a nonsafety-related load with the Division III emergency diesel generator at full power paralleled to the grid. To prevent isolating offsite power to any safety bus, the licensee issued procedure changes to prevent starting loads on the nonsafety-related buses connected to the divisional safety buses while the emergency diesel generators are in test. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-00910.
The finding was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Table 4a, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its 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, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. There is no crosscutting aspect associated with this violation because this is a historical condition not previously identified by the licensee.
Inspection Report# : 2010003 (pdf)
Significance:      Jun 02, 2010 Identified By: NRC Item Type: VIO Violation Failure to Ensure at Least One Train of Equipment Necessary to Achieve Hot Shutdown Conditions is Free of Fire Damage The team identified a cited violation of License Condition 2.C.(10), Fire Protection, for failing to ensure that the Division 1 standby service water support system to the Division 1 emergency diesel generator, which was required to achieve safe shutdown, was protected such that it remained free from fire damage under all conditions. This condition was identified by the licensee in May 2007, and entered into their corrective action program as a significant non-conforming condition in CR-RBS-2007-02102. The licensee subsequently initiated compensatory measures in the form of manual actions to protect the Division 1 emergency diesel generator. This issue was documented as a licensee-identified noncited violation in Inspection Report 2009002. River Bend has subsequently completed two refueling outages, six forced outages, and one emergency diesel generator work window of sufficient duration since identification of this condition and failed to correct the non-conformance. The team determined that schedule changes resulted in a new completion date of January 2011.
The failure to ensure that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) was free of fire damage and to correct this significant non-conforming condition in a timely manner is a performance deficiency. This performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A Phase 3 SDP risk assessment was performed by a senior reactor analyst. The bounding change in conditional core damage frequency for a 1-year exposure is the Fire Mitigation Frequency (4.30E-08/year)
 
multiplied by the change in conditional core damage probability (0.9) for a value of 3.87E-08/year. This value indicates the finding has very low safety significance (Green). Because the licensee failed to correct this violation, this violation is being treated as a cited violation, consistent with the NRC Enforcement Policy. This finding had a crosscutting aspect in the Work Control component of the Human Performance area because the licensee did not appropriately plan work activities to support long-term equipment reliability by limiting temporary modifications, operator workarounds, safety systems unavailability, and reliance on manual actions [H.3(b)]. (Section 1R05.01)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Alternative Shutdown Procedure Could be Implemented as Written The team identified a noncited violation of Technical Specification 5.4.1.d, Fire Protection Program Implementation. Specifically, Procedure AOP-0031 Shutdown from Outside the Main Control Room, Revision 307, had steps that could not be implemented as written. Two steps were to be performed before the necessary ac power was available, and two steps required diagnostic assessment without the availability of instrumentation.
The failure to ensure that Procedure AOP-0031, Revision 307 could be implemented as written is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Attachment 2 to Appendix F, Fire Protection Significance Determination Process, this issue was determined to be a safe shutdown finding, and was assigned a degradation rating of Low because the examples involved procedural deficiencies that could be compensated for by operator experience. Since this finding was assigned a low degradation rating, the safety significance screened as very low (Green). This finding was entered into the licensees corrective action program as CR-RBS-2010-01592, CR-RBS-2010-01831, CR-RBS-2010-01775, CR-RBS-2010-01821, and CR-RBS-2010-1846. This finding had a crosscutting aspect in the Resources component of the Human Performance area, in that the licensee did not ensure that procedures were complete, accurate, and up to date to assure nuclear safety [H.2.(c)]. (Section 1R05.05.b.1)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement and Maintain in Effect all Provisions of the Approved Fire Protection Program The team identified a noncited violation of License Condition 2.C.(10), Fire Protection, for the failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the team identified, during a timed walkdown of the procedure that it took operators over 6 minutes to isolate feedwater, but the simulator showed that the steam lines could be flooded in 2 minutes. Overfilling the reactor pressure vessel and flooding the main steam lines could make reactor core isolation cooling unavailable. Reactor core isolation cooling was credited for decay heat removal and inventory control in the event of a fire.
The failure to ensure that feedwater would be isolated prior to overfilling the reactor pressure vessel and flooding the main steam lines making reactor core isolation cooling unavailable is a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a control room fire that led to control room abandonment. The Phase 3 evaluation determined that the finding had very low safety significance because a fire in only one of 109 electrical cabinets in the control room could result in this overfill event. The finding was entered into
 
the licensees corrective action program as CR-RBS-2010-01808. The finding did not have a crosscutting aspect since it was not indicative of current performance, in that the licensee had established the incorrect response time more than three years prior to this finding. (Section 1R05.05.b.2)
Inspection Report# : 2010006 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement and Maintain in Effect all Provisions of the Approved Fire Protection Program The team identified a noncited violation of License Condition 2.C.(10), Fire Protection, related to the licensee's failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, during testing required by the approved fire protection program the licensee failed to adequately test the remote shutdown emergency transfer switch functions used to assure isolation of safe shutdown equipment from the control room in the event of a control room evacuation due to fire. The switch functions had not been adequately tested since 1997.
The failure to ensure isolation from the control room for safe shutdown equipment controlled from the remote shutdown panel during surveillance testing of emergency transfer switches is a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone in that it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown. Using Appendix F, , Degradation Rating Guidance Specific to Various Fire Protection Program Elements, the team determined that the finding constituted a low degradation of the safe shutdown area since the control room isolation feature was expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. This finding screened as having very low safety significance (Green). This violation was entered into the licensees corrective action program as CR-RBS-2010-01783. Because the emergency transfer switch surveillance procedures had been in effect since 1997, there was no crosscutting aspect associated with the violation, in that it is not indicative of current licensee performance. (Section 1R05.05.b.3)
Inspection Report# : 2010006 (pdf)
Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance:        Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit Instructions NRC inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1 for failure to follow radiation work permit instructions. Specifically, a team technician made an unauthorized entry into a posted high radiation area on a radiation work permit that did not grant access to that area. The licensee conducted a review of this event and issued a site-wide memorandum on procedural and management expectations associated with high radiation areas. The licensee entered the finding into the corrective action program as Condition Report CR RBS-2009-03953.
The failure to follow the instructions on a radiation work permit is a performance deficiency. The performance
 
deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone. It is associated with the exposure control attribute in that a worker not following radiation work permit instructions does not ensure adequate protection of the workers health and safety from additional/unintended personal exposure. Using the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Furthermore, the finding had an associated human performance crosscutting aspect in the work practices component because the worker did not use human error prevention techniques, such as self-checking [H.4(a)].
Inspection Report# : 2010003 (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 07, 2011
 
River Bend 1 2Q/2011 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding Failure to Plug a Main Condenser Tube in Accordance with an Approved Work Order The inspectors reviewed a self-revealing finding for the licensees failure to plug a main condenser tube in accordance with an approved work order. Specifically, a plastic tube plug was not replaced with the required brass plug causing a tube leak requiring the plant to reduce power. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-04526.
The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, in that the performance deficiency created a condition that upset plant stability by creating a condenser tube leak that prompted the plant to reduce power. The inspectors determined that the apparent cause of this finding was the licensees failure to use human performance error-prevention techniques to ensure that the tube plugging was performed correctly. This finding therefore has a crosscutting aspect in the work practices component of the human performance area because the licensee did not communicate and use human error prevention techniques commensurate with the risk of the assigned task, such that work activities are performed safely [(H.4(a)].
Inspection Report# : 2010005 (pdf)
Significance:      Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Licensed Operator Examination Integrity The inspectors identified a noncited violation of 10 CFR Part 55.49, Integrity of Examinations and Tests, for the failure of operations training personnel to ensure the integrity of an operating test administered to a licensed operations crew was maintained. One licensed operations crew received two scenarios for their operating test that had been previously administered to a licensed operations staff crew. This failure resulted in a compromise of examination integrity, but did not lead to an actual effect on the equitable and consistent administration of the examination. This finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions when adopting a 50 percent operating examination overlap practice
[H.1(b)].
The finding is more than minor because, if left uncorrected, the finding could have become more significant in that allowing untested licensed operators at the controls could be a precursor to a significant event if undetected performance deficiencies develop. The finding was determined to have very low safety significance (Green) because the finding resulted in a compromise of the integrity of operating test scenarios and compensatory actions were not immediately taken when the compromise should have been discovered. However, the equitable and consistent administration of the exam was not actually impacted by this compromise. The inspectors applied Inspection Manual Chapter 0609, Significance Determination Process, Appendix I, Licensed Operator Requalification Significance Determination Process, and determined that the finding should be dispositioned as a Green noncited violation.
Inspection Report# : 2010004 (pdf)
Mitigating Systems
 
Significance:        Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Failure to Track and Document Plant Equipment Oil Usage The inspectors identified a finding for the failure to properly document equipment oil additions in the oil lubrication accountability log per General Maintenance Procedure GMP-0015, Lubrication Procedure. To correct the programmatic deficiencies, the station revised General Maintenance Procedure GMP-0015 instructions to enhance and amplify the requirement to record all oil additions in the lubrication accountability log, revise preventative maintenance tasks that sample or change oil to explicitly state record oil additions in the lubrication accountability log, and to brief station personnel concerning changes to General Maintenance Procedure GMP-0015. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-02883.
The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency; did not represent either a loss of system safety function, an actual loss of safety function of a single train, or an actual loss of safety function; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the apparent cause of the performance deficiency was incomplete work package instructions that did not explicitly state to record oil additions in the lubrication accountability log per General Maintenance Procedure GMP-0015, thereby making equipment operability conclusions based on incomplete monitored trends suspect and potentially inaccurate. Consequently, this finding has a cross-cutting aspect in the human performance area associated with the resources component because the stations work packages lacked the necessary instructions to adequately control the lubrication monitoring program
[H.2(c)].
Inspection Report# : 2011003 (pdf)
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Develop a Preventive Maintenance Schedule to Specify Inspection or Replacement of the O-Ring in the High Pressure Core Spray Lower Motor Bearing Drain Plug The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1 for the licensees failure to determine the appropriate preventive maintenance strategy and task frequency for the o-ring in the high pressure core spray lower motor bearing drain plug. As immediate correction action, the licensee replaced the o-ring. At the conclusion of the inspection, the licensee was in the process of determining the appropriate replacement frequency.
The licensee entered this issue into their corrective action system as Condition Report CR-RBS-2010-05766.
This finding was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern, in that if the licensee did not develop a preventive maintenance schedule for periodically replacing the subject o-ring, degradation of that o-ring due to aging could allow a leak that would drain oil from the lower motor bearing and thus render the high pressure core spray pump inoperable. As described in Inspection Manual 0609 Appendix A, a Phase 2 analysis using the presolved worksheet determined that this finding had very low (Green) risk significance. This finding has a crosscutting aspect in the operating experience component of the problem identification & resolution area because the licensee did not systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience [P.2(a)].
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Two Examples of Completing Maintenance that Affected the Performance of Safety-Related Equipment but Was Not Properly Preplanned The inspectors reviewed a two-example self-revealing green noncited violation of Technical Specification 5.4.1 for two occasions on which the licensee completed maintenance that affected the performance of safety-related equipment
 
(high pressure core spray) but was not properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances As a result, the licensee overtorqued the high pressure core spray lower motor bearing drain plug causing the plug to fracture. This fracture resulted in excessive oil leakage that caused the pump to become inoperable. The violation is in the licensees corrective action program as Condition Report CR-RBS-2011-00224.
These performance deficiencies were more than minor and therefore constituted a finding because they affected the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. As described in Inspection Manual 0609 Appendix A, a Phase 2 analysis using the presolved worksheet determined that this finding had very low risk significance. The finding has a crosscutting aspect in the resources component of the human performance area because the apparent cause of the finding was a procedure that was not adequate to assure nuclear safety [H.2(c)].
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Required Quality Control Inspections Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. The licensee performed extensive reviews, and inspectors performed independent reviews of the licensees conclusions as well as independent sampling, to confirm that improper or missed inspections did not actually affect the operability of plant equipment. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective action program under Condition Reports CR-HQN-2009-01184 and CR-HQN-2010-0013.
The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the design control attribute of the Mitigating Systems Cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to have very low safety significance in Phase 1 of the significance determination process since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this performance deficiency involved a crosscutting aspect related to the human performance area associated with decision making [H.1 (a)] because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate.
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program.
Specifically, the individual assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as Condition Report CR-HQN-2010-00386.
 
Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems Cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no crosscutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago.
Inspection Report# : 2010005 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate High Pressure Core Spray Pump Room Cooler Bearing Maintenance A self-revealing, very low safety significance (Green) noncited violation 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was reviewed for the licensees failure to prescribe lubrication and installation of bearings on the high-pressure core spray room cooler motors by adequate procedures. In response to this finding, the licensee changed their procedure for performing material equivalency evaluations to require that, when plant components change and associated vendor-recommended maintenance schedules change, licensee personnel also update the corresponding preventive-maintenance tasks. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-02919.
This finding was more 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, in that this finding caused inoperability of the high-pressure core spray. The significance of this finding was determined by completing a Phase 3 analysis in accordance with Inspection Manual Chapter 0609, Appendix A, which determined that the incremental core damage probability maximum was 2x10-7, and that the finding was therefore of very low safety significance (Green). This finding did not represent current licensee performance and consequently did not have a cross-cutting aspect because the cause of this finding was that when the licensee replaced a component by a similar component from a different vendor, no licensee procedure required them to update the associated maintenance frequencies, and because before this finding was identified, the licensee had no reasonable opportunity to identify and correct that deficiency in that procedure.
Inspection Report# : 2010004 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: FIN Finding Inadequate Maintenance Results in Unplanned Opening of Main Turbine Bypass Valve A self-revealing finding of very low safety significance (Green) was identified when turbine bypass valve number 1 opened unexpectedly causing the reactor to exceed 100 percent core thermal power. Operators promptly lowered core thermal power to 90 percent to preserve margin to fuel thermal limits. A failed power supply and inadequate calibration and testing of the steam bypass and pressure regulation system and electro-hydraulic control system caused the event. Corrective actions include replacing system power supplies and revising applicable calibration and test instructions. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-03343.
This finding was more 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. Specifically, technical specification mitigation equipment (main turbine bypass system, end-of-cycle recirculation pump trip function, and rod block instrumentation functions) became inoperable. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that the finding was of very low safety significance (Green)
 
because the finding 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 this finding did not represent current licensee performance because the preventative maintenance schedule and calibration procedure were developed and approved over two years ago. Therefore, no crosscutting aspect was assigned to this finding.
Inspection Report# : 2010004 (pdf)
Significance:      Jun 02, 2010 Identified By: NRC Item Type: VIO Violation Failure to Ensure at Least One Train of Equipment Necessary to Achieve Hot Shutdown Conditions is Free of Fire Damage The team identified a cited violation of License Condition 2.C.(10), Fire Protection, for failing to ensure that the Division 1 standby service water support system to the Division 1 emergency diesel generator, which was required to achieve safe shutdown, was protected such that it remained free from fire damage under all conditions. This condition was identified by the licensee in May 2007, and entered into their corrective action program as a significant non-conforming condition in CR-RBS-2007-02102. The licensee subsequently initiated compensatory measures in the form of manual actions to protect the Division 1 emergency diesel generator. This issue was documented as a licensee-identified noncited violation in Inspection Report 2009002. River Bend has subsequently completed two refueling outages, six forced outages, and one emergency diesel generator work window of sufficient duration since identification of this condition and failed to correct the non-conformance. The team determined that schedule changes resulted in a new completion date of January 2011.
The failure to ensure that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) was free of fire damage and to correct this significant non-conforming condition in a timely manner is a performance deficiency. This performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A Phase 3 SDP risk assessment was performed by a senior reactor analyst. The bounding change in conditional core damage frequency for a 1-year exposure is the Fire Mitigation Frequency (4.30E-08/year) multiplied by the change in conditional core damage probability (0.9) for a value of 3.87E-08/year. This value indicates the finding has very low safety significance (Green). Because the licensee failed to correct this violation, this violation is being treated as a cited violation, consistent with the NRC Enforcement Policy. This finding had a crosscutting aspect in the Work Control component of the Human Performance area because the licensee did not appropriately plan work activities to support long-term equipment reliability by limiting temporary modifications, operator workarounds, safety systems unavailability, and reliance on manual actions [H.3(b)]. (Section 1R05.01)
Inspection Report# : 2010006 (pdf)
Barrier Integrity 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 : October 14, 2011
 
River Bend 1 3Q/2011 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2011 Identified By: NRC Item Type: FIN Finding Ineffective Corrective Actions on the Main Steam Equalizing Header Drain Bypass Valve Results in an Unplanned Down Power The inspectors identified a self-revealing finding involving inadequate corrective actions in response to a failure in the main steam equalizing header drain bypass valve, resulting in a steam leak and an unplanned plant down power.
Specifically, plant personnel failed to properly address the dual indication on the bypass valve and fluid flow through the valve caused water to flash to steam accelerating pipe wall erosion and piping failure. The licensees immediate corrective actions were to identify, secure, and temporarily repair the steam leak. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-04592.
The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it was not a loss of coolant accident initiator, did not contribute to both the likelihood of an initiating event and the likelihood that mitigating equipment or functions would not be available, nor increase the likelihood of an external event (seismic, flooding, or severe weather event). The apparent cause of the performance deficiency was that the control room and outage control center personnel presumed that the main control room dual indication for the valve was incorrect because previously valve operation successfully closed the valve. Consequently, this finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel did not use a systematic process to assess the condition of the bypass valve, and failed to verify the validity of the underlying assumptions that were used to justify operation with the valve having dual indications [H.1(a)].
Inspection Report# : 2011004 (pdf)
Significance:        Mar 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Feedwater Control System Inadequate Corrective Actions Results in Power Transient The inspectors reviewed a self-revealing finding involving failure to take adequate corrective actions on a degraded feedwater flow controller push-button, causing a recirculation flow control valve runback, reactor vessel level transient, and a resulting reactor power transient. On September 24, 2008, operations documented a deficiency in the function of the push-button, however station maintenance personnel failed to adequately address the identified deficiency. The push button was subsequently repaired and this issue was entered into the licensees corrective action program as Condition Report CR-RBS-2011-00300.
The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone, and it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined it to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined that the apparent cause of the performance deficiency was the
 
failure to thoroughly evaluate the cause of the defective push-buttons stickiness. Consequently, 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 adequately review the results of the work order to ensure that the cause and extent of condition of the defective push-button was resolved in a timely manner [P.1(c)].
Inspection Report# : 2011002 (pdf)
Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding Failure to Plug a Main Condenser Tube in Accordance with an Approved Work Order The inspectors reviewed a self-revealing finding for the licensees failure to plug a main condenser tube in accordance with an approved work order. Specifically, a plastic tube plug was not replaced with the required brass plug causing a tube leak requiring the plant to reduce power. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2010-04526.
The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, in that the performance deficiency created a condition that upset plant stability by creating a condenser tube leak that prompted the plant to reduce power. The inspectors determined that the apparent cause of this finding was the licensees failure to use human performance error-prevention techniques to ensure that the tube plugging was performed correctly. This finding therefore has a crosscutting aspect in the work practices component of the human performance area because the licensee did not communicate and use human error prevention techniques commensurate with the risk of the assigned task, such that work activities are performed safely [(H.4(a)].
Inspection Report# : 2010005 (pdf)
Mitigating Systems Significance:      Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Failure to Track and Document Plant Equipment Oil Usage The inspectors identified a finding for the failure to properly document equipment oil additions in the oil lubrication accountability log per General Maintenance Procedure GMP-0015, Lubrication Procedure. To correct the programmatic deficiencies, the station revised General Maintenance Procedure GMP-0015 instructions to enhance and amplify the requirement to record all oil additions in the lubrication accountability log, revise preventative maintenance tasks that sample or change oil to explicitly state record oil additions in the lubrication accountability log, and to brief station personnel concerning changes to General Maintenance Procedure GMP-0015. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-02883.
The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency; did not represent either a loss of system safety function, an actual loss of safety function of a single train, or an actual loss of safety function; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the apparent cause of the performance deficiency was incomplete work package instructions that did not explicitly state to record oil additions in the lubrication accountability log per General Maintenance Procedure GMP-0015, thereby making equipment operability conclusions based on incomplete monitored trends suspect and potentially inaccurate. Consequently, this finding has a cross-cutting aspect in the human performance area associated with the resources component because the stations work packages lacked the necessary instructions to adequately control the lubrication monitoring program
[H.2(c)].
Inspection Report# : 2011003 (pdf)
 
Significance:        May 12, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Take Corrective Action for Service-Induced Failures of Gould J-series Relays The inspectors reviewed a self-revealing green noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take corrective action to address service-induced failures of Gould J series relays. In response, the licensee initiated condition report CR RBS 2010 06032 to ensure that appropriate levels of preventive maintenance are performed on high-critical components.
The performance deficiency was the licensee's failure to take adequate corrective actions to address service-induced failures of the high-critical, high-duty-cycle Gould J series relay designated as EHS MCC16B6D 33X1. This performance deficiency was determined to be more than minor and was therefore a finding because it impacted the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding had very low safety significance because the finding was not a design or qualification deficiency confirmed not to result in a loss of operability, did not represent a loss of system safety function, did not represent 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. Because the apparent cause of this finding was the licensees misclassification of the failed relay within the preventive maintenance optimization program in 2008, and because the licensees performance in that program was not reflective of current licensee performance, no cross-cutting aspect was assigned to this finding.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Repetitive Service Water Pressure Control Valves Diaphragm Failures Affecting Control Building Chillers Operability The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly identify and correct adverse service water chemistry conditions to resolve repetitive service water pressure control valves diaphragm failures that affected operability of the control building chillers. Specifically, station personnel failed to address excessive internal corrosion in the pressure control valves, which resulted in loss of service water pressure control to the control building chillers. As immediate corrective action, the licensee replaced the damaged pressure control valve and is currently evaluating methods to preclude corrosion around the diaphragm.
The licensee placed this issue into their corrective action program as Condition Report CR-RBS-2011-02126.
The finding was more than minor because it was associated with the equipment performance attribute of the reactor safety Mitigating Systems (MS) 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 Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green), because it did not result in a loss of system safety function. The inspectors determined that the apparent cause of the performance deficiency was the repetitive failure of 1SWP-PVY32 diaphragm from rust barnacles that formed on the valve internal steel parts during low flow conditions. The apparent cause of the performance deficiency was the stations failure to thoroughly evaluate the cause of the corrosion build up mechanism because the station treated diaphragm failures as a broke/fix maintenance item. Consequently, this finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to minimize long-standing equipment issues [H.2(a)].
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Determine the Appropriate Preventive Maintenance Strategy and Task Frequency for the Reactor Core Isolation Cooling System Turbine Lube Oil Cooler Inlet Pressure Control Valve
 
The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1 for the licensees failure to determine the appropriate preventive maintenance strategy and task frequency for the reactor core isolation cooling system turbine lube oil cooler inlet pressure control valve (E51-PCVF015). The vendor manual for the pressure control valve recommends that non-metallic parts (including diaphragms) be replaced after 5 years in service. On October 13, 2010, after being in service for more than ten years without diaphragm replacement, the valve developed a leak that rendered the reactor core isolation cooling system inoperable. The licensee replaced the damaged diaphragm and created a preventive maintenance activity for its periodically replacement. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2010-05224.
This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the subject performance deficiency allowed a failure to occur that rendered the reactor core isolation cooling system inoperable for approximately 14 hours. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, to assess its risk significance. The reactor core isolation cooling diaphragm failure was determined to have occurred when the pump was secured; that is, the pump could have operated for 24 hours if it had not been shut down at that time. Therefore, the exposure time was equal to the repair time, which was 15.5 hours. The finding involved a loss of safety system function and therefore did not screen in Phase 1, requiring a Phase 2 evaluation. The inspectors used the Phase 2 pre-solved spreadsheet with a duration of 0-3 days to determine that the issue had very low significance (Green). The inspectors concurred with the licensees determination that a lack of technical rigor had been the reason why the preventive maintenance evaluation of valve E51-PCVF015 had been incorrect, and was therefore the major contributor to the finding. The inspectors considered that this contributor does not reflect current licensee performance because this contributor is a human performance error that occurred in September 2006, and because in 2007, the licensee developed corrective actions to address a substantive crosscutting issue in human performance. Those actions are described in Condition Report CR-RBS-2007-00835 and included activities that changed the licensees human performance program such that the human performance error that occurred in September of 2006 is not likely to re-occur. This finding therefore does not have a crosscutting aspect.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control Rod Inspection Procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving an inadequate control rod inspection procedure. Specifically, the stations procedures only required inspection of a only 20 percent of the control rods that exceeded the inspection criteria, instead of all of them.
The station currently has 18 CR 82M control rods in the reactor core in shutdown locations that have exceeded Westinghouses inspection threshold exposure limits. In response to the inspectors inquries, the licensee reviewed their water chemistry and concluded the current tritium and boron levels indicated there was margin for control rod operability. The licensee intends to monitor the reactor coolant for increasing boron and tritium levels throughout this operating cycle. The licensee placed this issue into their corrective action program as Condition Report CR-RBS-2011-01704.
The finding is more than minor because it is associated with the equipment performance attribute of the reactor safety 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 Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to be of very low safety significance (Green) because it did not result in a loss of system safety function. The inspectors determined that the apparent cause of the performance deficiency was River Bend Stations failure to communicate relevant operating experience to affected internal and external stakeholders. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the operating experience component because the licensee failed to appropriately apply all the CR 82M control rod inspection requirements provided by the control rod vendor [P.2(b)].
Inspection Report# : 2011002 (pdf)
 
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Develop a Preventive Maintenance Schedule to Specify Inspection or Replacement of the O-Ring in the High Pressure Core Spray Lower Motor Bearing Drain Plug The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1 for the licensees failure to determine the appropriate preventive maintenance strategy and task frequency for the o-ring in the high pressure core spray lower motor bearing drain plug. As immediate correction action, the licensee replaced the o-ring. At the conclusion of the inspection, the licensee was in the process of determining the appropriate replacement frequency.
The licensee entered this issue into their corrective action system as Condition Report CR-RBS-2010-05766.
This finding was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern, in that if the licensee did not develop a preventive maintenance schedule for periodically replacing the subject o-ring, degradation of that o-ring due to aging could allow a leak that would drain oil from the lower motor bearing and thus render the high pressure core spray pump inoperable. As described in Inspection Manual 0609 Appendix A, a Phase 2 analysis using the presolved worksheet determined that this finding had very low (Green) risk significance. This finding has a crosscutting aspect in the operating experience component of the problem identification & resolution area because the licensee did not systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience [P.2(a)].
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Two Examples of Completing Maintenance that Affected the Performance of Safety-Related Equipment but Was Not Properly Preplanned The inspectors reviewed a two-example self-revealing green noncited violation of Technical Specification 5.4.1 for two occasions on which the licensee completed maintenance that affected the performance of safety-related equipment (high pressure core spray) but was not properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances As a result, the licensee overtorqued the high pressure core spray lower motor bearing drain plug causing the plug to fracture. This fracture resulted in excessive oil leakage that caused the pump to become inoperable. The violation is in the licensees corrective action program as Condition Report CR-RBS-2011-00224.
These performance deficiencies were more than minor and therefore constituted a finding because they affected the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. As described in Inspection Manual 0609 Appendix A, a Phase 2 analysis using the presolved worksheet determined that this finding had very low risk significance. The finding has a crosscutting aspect in the resources component of the human performance area because the apparent cause of the finding was a procedure that was not adequate to assure nuclear safety [H.2(c)].
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Required Quality Control Inspections Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. The licensee performed extensive reviews, and inspectors performed independent reviews of the licensees conclusions as well as independent sampling, to confirm that improper or missed inspections did not actually affect the operability of plant equipment. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective
 
action program under Condition Reports CR-HQN-2009-01184 and CR-HQN-2010-0013.
The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the design control attribute of the Mitigating Systems Cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to have very low safety significance in Phase 1 of the significance determination process since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this performance deficiency involved a crosscutting aspect related to the human performance area associated with decision making [H.1 (a)] because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate.
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program.
Specifically, the individual assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as Condition Report CR-HQN-2010-00386.
Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems Cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no crosscutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago.
Inspection Report# : 2010005 (pdf)
Significance:        Jun 02, 2010 Identified By: NRC Item Type: VIO Violation Failure to Ensure at Least One Train of Equipment Necessary to Achieve Hot Shutdown Conditions is Free of Fire Damage The team identified a cited violation of License Condition 2.C.(10), Fire Protection, for failing to ensure that the Division 1 standby service water support system to the Division 1 emergency diesel generator, which was required to achieve safe shutdown, was protected such that it remained free from fire damage under all conditions. This condition was identified by the licensee in May 2007, and entered into their corrective action program as a significant non-conforming condition in CR-RBS-2007-02102. The licensee subsequently initiated compensatory measures in the form of manual actions to protect the Division 1 emergency diesel generator. This issue was documented as a licensee-identified noncited violation in Inspection Report 2009002. River Bend has subsequently completed two refueling outages, six forced outages, and one emergency diesel generator work window of sufficient duration since identification of this condition and failed to correct the non-conformance. The team determined that schedule changes
 
resulted in a new completion date of January 2011.
The failure to ensure that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) was free of fire damage and to correct this significant non-conforming condition in a timely manner is a performance deficiency. This performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post fire safe shutdown systems with plant-wide consequences. A Phase 3 SDP risk assessment was performed by a senior reactor analyst. The bounding change in conditional core damage frequency for a 1-year exposure is the Fire Mitigation Frequency (4.30E-08/year) multiplied by the change in conditional core damage probability (0.9) for a value of 3.87E-08/year. This value indicates the finding has very low safety significance (Green). Because the licensee failed to correct this violation, this violation is being treated as a cited violation, consistent with the NRC Enforcement Policy. This finding had a crosscutting aspect in the Work Control component of the Human Performance area because the licensee did not appropriately plan work activities to support long-term equipment reliability by limiting temporary modifications, operator workarounds, safety systems unavailability, and reliance on manual actions [H.3(b)]. (Section 1R05.01)
Inspection Report# : 2010006 (pdf)
Barrier Integrity Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Standby Gas Treatment Electric Heater Power Output Calculation The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, for an inadequate calculation methodology used in determining standby gas treatment system operability. The inspectors found that the calculation neither considered instrument uncertainty nor applied a proper voltage drop from the breaker to the standby gas treatment system filter train heater. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-07332.
The finding was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone to maintain radiological barrier functionality of standby gas treatment trains, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, operating the standby gas system filter train heaters without sufficient output power is detrimental to the charcoal filters ability to retain radioactive iodine. This could result in a greater amount of radiation release to the environment in the event of an accident. In accordance with Inspection manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the Phase 1 significance determination process screening determined the finding to be only of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the standby gas treatment system. The apparent cause of this finding was the decision to develop an engineering evaluation that did not include instrument uncertainly and did not validate the correct voltage drop between the filter train heater feeder breaker and the heater elements. The finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel failed to use conservative assumptions when developing the modified output power methodology for operation of the standby gas treatment system filter heaters with only 8 of 9 heater elements installed [H.1(b)].
Inspection Report# : 2011004 (pdf)
Significance:      Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation
 
Failure to Implement Procedure AOP-0027, "Fuel Handling Mishaps" The inspector identified a Green noncited violation of Technical Specification 5.4.1.a, Procedures for River Bend Station fuel handling personnel failing to follow AOP-0027, Fuel Handling Mishaps, when an actual fuel handling event occurred. Instead of entering the AOP, fuel handling personnel continued to move a fuel assembly after equipment damage and potential fuel damage. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-03692.
This failure to follow procedures is a performance deficiency. The performance deficiency is more that minor, and therefore a finding, because it adversely impacted the human performance attribute of the barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, the inspector determined the finding had very low safety significance (Green) because the fuel cladding barrier was potentially degraded but there was no release of radionuclides. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the operating experience component because the licensee failed to implement and institutionalize operating experience through changes to station procedures and training programs [P.2(b)].
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: FIN Finding Failure to Follow Fuel Handling Guidelines The inspector identified a finding for failure to follow River Bend Stations Fuel Handling Guideline. A fuel handling event occurred at River Bend Station on January 21, 2011, when a fuel assembly was grappled and raised approximately one foot rather than fully withdrawn from the core. With the fuel assembly only partially withdrawn from the core, the refuel platform was erroneously moved horizontally approximately five feet. This inappropriate stop at one foot followed by inappropriate horizontal movement of the refuel platform with the fuel partially inserted into the core resulted in equipment damage and potential fuel damage. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-03693.
This failure to follow the guideline is a performance deficiency. The performance deficiency is more that minor, and therefore a finding, because it adversely impacted the human performance attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, the inspector determined the finding had very low safety significance (Green) because the fuel cladding barrier was potentially degraded but there was no release of radionuclides. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee made a safety-significant decision without verifying the validity of underlying assumptions [H.1(b)].
Inspection Report# : 2011002 (pdf)
Emergency Preparedness Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Reduction in ERO Staffing Decreased Emergency Plan Effectiveness The inspector identified a Severity Level IV noncited violation of 10 CFR 50.54(q) for changes to the licensees emergency plans that decreased the effectiveness of those plans without NRC approval. Specifically, the effectiveness of River Bend Station Emergency Plan, Revision 36, was reduced by removal of the Health Physics Communicator position from the emergency response organization. The licensees failure to recognize that Revision 36 decreased the effectiveness of licensee emergency plans was a performance deficiency. The licensee has entered this issue into their corrective action system as CR-RBS-2011-02366.
 
This finding is more than minor because it has a potential effect on the licensees emergency response capabilities and because the licensee may not be capable of implementing adequate measures to protect the health and safety of the public when the effectiveness of its emergency response organization has been reduced. The finding was evaluated using the NRC Enforcement Policy because it impeded the regulatory process as defined by Manual Chapter 0609, Appendix B, Section 2.2(e). The finding was determined to be Severity Level IV because it decreased the licensees ability to meet or implement a regulatory requirement not related to assessment or notification.
Inspection Report# : 2011002 (pdf)
Occupational Radiation Safety Significance:      Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Control and Guard the Entrance to a Locked High Radiation Area The inspectors identified a noncited violation of Technical Specification 5.7.2 for failure to properly control and guard a high radiation area with dose rates greater than or equal to 1000 mrem/hr. Specifically, on January 25, 2011, while touring the outside area between the auxiliary building and the radioactive waste building, the inspectors noted that the access gate to a locked high radiation area was open. A guard for the locked high radiation area was positioned in a tent enclosure to the right of the gate, but was not in a position to maintain line-of-sight control of the access to the locked high radiation area. The licensee immediately repositioned the guard and enhanced the tent construction to provide the necessary control for access to the area. The licensee placed this issue into their corrective action program as CR 2011-01154.
The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to properly control access to a high radiation area with dose rates in excess of 1000 mrem/hr had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because it was not associated with ALARA planning or work controls, there was no overexposure, there was no substantial potential for an overexposure, and the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work control, work planning activities, because the individuals failed to consider job site conditions which would impact the ability of the guard to adequately observe the entrance to the locked high radiation area [H.3(a)].
Inspection Report# : 2011002 (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 Significance: N/A May 12, 2011 Identified By: NRC
 
Item Type: FIN Finding River Bend Plant Biennial PI&R Inspection Summary The team determined that the licensees program for identifying, prioritizing, and correcting conditions adverse to quality was effective. With few exceptions, the licensee identified conditions adverse to quality at a low threshold, properly classified and evaluated those conditions, and developed appropriate corrective actions.
The licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items in the corrective action program. The licensee used industry operating experience when performing root cause and apparent cause evaluations. In addition, the licensee performed effective quality assurance audits and self-assessments.
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# : 2011006 (pdf)
Last modified : January 04, 2012
 
River Bend 1 4Q/2011 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2011 Identified By: NRC Item Type: FIN Finding Ineffective Corrective Actions on the Main Steam Equalizing Header Drain Bypass Valve Results in an Unplanned Down Power The inspectors identified a self-revealing finding involving inadequate corrective actions in response to a failure in the main steam equalizing header drain bypass valve, resulting in a steam leak and an unplanned plant down power.
Specifically, plant personnel failed to properly address the dual indication on the bypass valve and fluid flow through the valve caused water to flash to steam accelerating pipe wall erosion and piping failure. The licensees immediate corrective actions were to identify, secure, and temporarily repair the steam leak. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-04592.
The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it was not a loss of coolant accident initiator, did not contribute to both the likelihood of an initiating event and the likelihood that mitigating equipment or functions would not be available, nor increase the likelihood of an external event (seismic, flooding, or severe weather event). The apparent cause of the performance deficiency was that the control room and outage control center personnel presumed that the main control room dual indication for the valve was incorrect because previously valve operation successfully closed the valve. Consequently, this finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel did not use a systematic process to assess the condition of the bypass valve, and failed to verify the validity of the underlying assumptions that were used to justify operation with the valve having dual indications [H.1(a)].
Inspection Report# : 2011004 (pdf)
Significance:        Mar 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Feedwater Control System Inadequate Corrective Actions Results in Power Transient The inspectors reviewed a self-revealing finding involving failure to take adequate corrective actions on a degraded feedwater flow controller push-button, causing a recirculation flow control valve runback, reactor vessel level transient, and a resulting reactor power transient. On September 24, 2008, operations documented a deficiency in the function of the push-button, however station maintenance personnel failed to adequately address the identified deficiency. The push button was subsequently repaired and this issue was entered into the licensees corrective action program as Condition Report CR-RBS-2011-00300.
The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone, and it affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Phase 1 of Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and determined it to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined that the apparent cause of the performance deficiency was the
 
failure to thoroughly evaluate the cause of the defective push-buttons stickiness. Consequently, 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 adequately review the results of the work order to ensure that the cause and extent of condition of the defective push-button was resolved in a timely manner [P.1(c)].
Inspection Report# : 2011002 (pdf)
Mitigating Systems Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Testing of Division I and Division III Standby Diesel Generators The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, prior to October 27, 2011, the licensee failed to ensure surveillance testing procedures of Division I and III standby diesel generators incorporated the correct acceptance limits for maximum expected load at max frequency and voltage specified in design basis documents. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07132, CR-RBS-2011-07294, and CR-RBS-2011-07518.
The team determined that the failure to ensure that the test procedures required to demonstrate that Division I and Division III standby diesel generators will perform satisfactorily in service incorporated the requirements and acceptance limits contained in applicable design documents was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee could not ensure that the standby diesel generators would reliably provide power for the maximum expected post-accident loads including maximum frequency and voltage. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1
- Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of condition
[P.1(c)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Conservative Design Assumptions in the Ultimate Heat Sink Inventory Calculation The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 27, 2011, the licensee failed to assure that the design basis information for expected heat loads to the ultimate heat sink was correctly translated into the ultimate heat sink 30-day inventory analysis. The analysis used a less conservative, frictionless form of the conservation of energy equation to determine heat load in the standby service water system during a 30-day design basis event. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07430 and CR-RBS-2011-07654.
The team determined that the failure to correctly translate expected heat loads into the ultimate heat sink inventory
 
analysis was a performance deficiency. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to undesired consequences. Specifically, the neglect of friction heat load in the ultimate heat sink analysis system resulted in a condition where there was reasonable doubt on the operability of a system to meet its 30-day mission time without a makeup water source. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Specifically, the licensees revised analysis to determine operability removed overly conservative assumptions for operating the low pressure core spray pump for 30 days to account for the friction heat load added to the system. The finding has a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to thoroughly evaluate problems such that the resolutions address cause and extent of condition [P.1(c)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Residual Heat Removal Heat Exchanger Testing Frequency The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, from October 1998 to October 27, 2011, the licensee failed to establish a NRC Generic Letter 89-13 test program which incorporated a final test frequency for the residual heat removal heat exchangers and perform an adequate trending analysis upon which to base a final test frequency. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07713.
The team determined that failure to establish a NRC Generic Letter 89-13 test program which incorporated a final testing frequency of the residual heat removal heat exchangers was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inappropriate test frequency affected the licensees ability to ensure residual heat removal heat exchangers, when called upon, were available and capable to reliably perform as expected. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determine to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding did not have a crosscutting aspect because the most significance contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Monitoring Standby Service Water System Leakage The team identified a Green, noncited violation of 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, which states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
Specifically, prior to October 27, 2011, the licensee failed to provide appropriate quantitative or qualitative acceptance criteria in station and abnormal operating procedures to determine if actions for leak detection were satisfactorily accomplished to protect the standby service water system and ultimate heat sink during design basis events. This
 
finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07555.
The team determined that the failure to include appropriate acceptance criteria for leak detection in abnormal operating procedures for the standby service water system and ultimate heat sink was a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inadequate procedure guidance could lead to operators not recognizing conditions that would degrade the availability of the standby service water system. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 -
Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance: SL-IV Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for Change to Ultimate Heat Sink Inventory Requirements The team identified a Severity Level IV, noncited violation of 10 CFR 50.59, Changes, Tests and Experiments which states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if this activity would; result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the final safety analysis report (as updated). Specifically, from December 16, 2002, to October 27, 2011, the licensee changed the design basis of the ultimate heat sink inventory requirements to provide a 30-day cooling water supply without makeup capability to providing a less than 30-day cooling water supply with makeup capability without obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR 2011-07674.
The team determined that the failure to obtain a license amendment prior to implementing a proposed change, test or experiment to the ultimate heat sink requirements was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding has the ability to impact the regulatory process. The finding was more than minor because it involved a change to the updated final safety analysis report description where there was a reasonable likelihood that the change would require NRC approval. In accordance with the NRC Enforcement Policy, the team used insights from MC 0609, Significance Determination Process, to determine the final significance of the finding. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 -
Initial Screening and Characterization of Findings, the finding represented a loss of system safety function in that the ultimate heat sink could not meet its 30-day mission time to provide decay heat removal. Therefore, a Phase 2 evaluation was necessary. The significance of the finding could not be assessed quantitatively through a Phase 2 or Phase 3 analysis. Consequently, an assessment was performed in accordance with IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The finding was determined to have very low safety significance because the frequency of events that would require long term use of the ultimate heat sink is very low and the difference in the failure probability to replenish the ultimate heat sink in 10 days versus 30 days is very small. This was because an early depletion of the inventory would be easily detected and would become a priority. At the time that replenishment would be needed, plant conditions should be stable and local transportation arteries should be restored. Therefore, since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance:      Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Abnormal Procedure for Reducing Loads on Standby Diesel Generators
 
The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, and drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, prior to October 27, 2011, the licensee failed to include appropriate qualitative and quantitative acceptance criteria in abnormal operating procedures for control room operators to recognize the need to reduce loads on the standby diesel generators during design basis accidents. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07716.
The team determined that the failure to include appropriate quantitative or qualitative acceptance criteria in abnormal operating procedures for control room operators to recognize the need to reduce loads on the standby diesel generators during design basis accidents was a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, a control room operating crews failure to recognize the need to reduce loads to prevent the standby diesel generator failure during design basis accidents adversely affected the reliability of the standby diesel generators. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety for the correct training of licensed operator personnel [H.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency and Abnormal Procedures for Standby Diesel Generator Fail to Load Sequences The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, and drawings shall include appropriate qualitative and quantitative criteria for determining that important activities have been satisfactorily accomplished.
Specifically, prior to October 27, 2011, the licensee failed to include appropriate quantitative or qualitative acceptance criteria in procedures for control room operators to recognize and recover a standby diesel generator that starts but fails to load with the remaining standby diesel generator out of service during a loss-of-offsite-power event. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07716, CR-RBS-2011-07717, and CR-RBS-2011-07718.
The team determined that the failure to include appropriate quantitative or qualitative acceptance criteria to determine that important activities are satisfactorily accomplished in emergency and abnormal operating procedures used during loss-of-offsite-power events was a performance deficiency. The finding was more than minor because it is associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, a control room operator crews failure to diagnose recoverable conditions adversely affected the availability of standby diesel generators during a loss-of-offsite-power event. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of problem identification and resolution, operating experience component, because the licensee did not implement and institutionalize operating experience through changes to station processes, procedures, equipment, and training programs [P.2(b)].
Inspection Report# : 2011008 (pdf)
 
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Simulator Fidelity for Emergency Diesel Generator Loading The team identified a Green, noncited violation of 10 CFR 55.46(c)(1), Simulation Facilities, which states, in part, that a plant-referenced simulator used for the administration of the operating test must demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond. Specifically, prior to October 27, 2011, the River Bend Station simulator did not demonstrate the expected plant response for standby diesel generator loading during accident conditions to which the simulator was designed to respond. The electrical loading on the emergency diesel generator in the simulator was approximately 800 kW less than the expected full load for the diesel generator. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07682.
The team determined that the failure of the plant-referenced simulator to demonstrate expected plant response for standby diesel generator loading during accident conditions to which the simulator has been designed to respond was a performance deficiency. The finding was more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the incorrect simulator response adversely affected the control room operator crews capability to assess standby diesel generator loading conditions. In accordance with NRC Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets and the associated Appendix I, the finding was determined to be of very low safety significance (Green). Specifically, Manual Chapter 0609, Appendix I, Operator Requalification Human Performance Significance Determination Process, block 12, establishes a Green finding for failure to correctly replicate the plants response on the simulator that either has the potential to cause or actually causes negative training to operators. Negative training did occur for this finding because operators thought they had electrical load margin on the emergency diesel generators when the diesels were actually fully loaded with minimal margin without securing other equipment. This finding had a crosscutting aspect in the area of human performance, resources component, in that the licensee did not ensure that equipment (plant-referenced simulator) was adequate to assure nuclear safety for the correct training of licensed operator personnel [H.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Failure to Track and Document Plant Equipment Oil Usage The inspectors identified a finding for the failure to properly document equipment oil additions in the oil lubrication accountability log per General Maintenance Procedure GMP-0015, Lubrication Procedure. To correct the programmatic deficiencies, the station revised General Maintenance Procedure GMP-0015 instructions to enhance and amplify the requirement to record all oil additions in the lubrication accountability log, revise preventative maintenance tasks that sample or change oil to explicitly state record oil additions in the lubrication accountability log, and to brief station personnel concerning changes to General Maintenance Procedure GMP-0015. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-02883.
The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency; did not represent either a loss of system safety function, an actual loss of safety function of a single train, or an actual loss of safety function; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the apparent cause of the performance deficiency was incomplete work package instructions that did not explicitly state to record oil additions in the lubrication accountability log per General Maintenance Procedure GMP-0015, thereby making equipment operability conclusions based on incomplete monitored trends suspect and potentially inaccurate. Consequently, this finding has a cross-cutting aspect in the human performance area associated with the resources component because the stations work packages lacked the necessary instructions to adequately control the lubrication monitoring program
[H.2(c)].
 
Inspection Report# : 2011003 (pdf)
Significance:        May 12, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Take Corrective Action for Service-Induced Failures of Gould J-series Relays The inspectors reviewed a self-revealing green noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take corrective action to address service-induced failures of Gould J series relays. In response, the licensee initiated condition report CR RBS 2010 06032 to ensure that appropriate levels of preventive maintenance are performed on high-critical components.
The performance deficiency was the licensee's failure to take adequate corrective actions to address service-induced failures of the high-critical, high-duty-cycle Gould J series relay designated as EHS MCC16B6D 33X1. This performance deficiency was determined to be more than minor and was therefore a finding because it impacted the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding had very low safety significance because the finding was not a design or qualification deficiency confirmed not to result in a loss of operability, did not represent a loss of system safety function, did not represent 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. Because the apparent cause of this finding was the licensees misclassification of the failed relay within the preventive maintenance optimization program in 2008, and because the licensees performance in that program was not reflective of current licensee performance, no cross-cutting aspect was assigned to this finding.
Inspection Report# : 2011006 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Repetitive Service Water Pressure Control Valves Diaphragm Failures Affecting Control Building Chillers Operability The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly identify and correct adverse service water chemistry conditions to resolve repetitive service water pressure control valves diaphragm failures that affected operability of the control building chillers. Specifically, station personnel failed to address excessive internal corrosion in the pressure control valves, which resulted in loss of service water pressure control to the control building chillers. As immediate corrective action, the licensee replaced the damaged pressure control valve and is currently evaluating methods to preclude corrosion around the diaphragm.
The licensee placed this issue into their corrective action program as Condition Report CR-RBS-2011-02126.
The finding was more than minor because it was associated with the equipment performance attribute of the reactor safety Mitigating Systems (MS) 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 Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green), because it did not result in a loss of system safety function. The inspectors determined that the apparent cause of the performance deficiency was the repetitive failure of 1SWP-PVY32 diaphragm from rust barnacles that formed on the valve internal steel parts during low flow conditions. The apparent cause of the performance deficiency was the stations failure to thoroughly evaluate the cause of the corrosion build up mechanism because the station treated diaphragm failures as a broke/fix maintenance item. Consequently, this finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to minimize long-standing equipment issues [H.2(a)].
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation
 
Failure to Determine the Appropriate Preventive Maintenance Strategy and Task Frequency for the Reactor Core Isolation Cooling System Turbine Lube Oil Cooler Inlet Pressure Control Valve The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1 for the licensees failure to determine the appropriate preventive maintenance strategy and task frequency for the reactor core isolation cooling system turbine lube oil cooler inlet pressure control valve (E51-PCVF015). The vendor manual for the pressure control valve recommends that non-metallic parts (including diaphragms) be replaced after 5 years in service. On October 13, 2010, after being in service for more than ten years without diaphragm replacement, the valve developed a leak that rendered the reactor core isolation cooling system inoperable. The licensee replaced the damaged diaphragm and created a preventive maintenance activity for its periodically replacement. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2010-05224.
This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the subject performance deficiency allowed a failure to occur that rendered the reactor core isolation cooling system inoperable for approximately 14 hours. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, to assess its risk significance. The reactor core isolation cooling diaphragm failure was determined to have occurred when the pump was secured; that is, the pump could have operated for 24 hours if it had not been shut down at that time. Therefore, the exposure time was equal to the repair time, which was 15.5 hours. The finding involved a loss of safety system function and therefore did not screen in Phase 1, requiring a Phase 2 evaluation. The inspectors used the Phase 2 pre-solved spreadsheet with a duration of 0-3 days to determine that the issue had very low significance (Green). The inspectors concurred with the licensees determination that a lack of technical rigor had been the reason why the preventive maintenance evaluation of valve E51-PCVF015 had been incorrect, and was therefore the major contributor to the finding. The inspectors considered that this contributor does not reflect current licensee performance because this contributor is a human performance error that occurred in September 2006, and because in 2007, the licensee developed corrective actions to address a substantive crosscutting issue in human performance. Those actions are described in Condition Report CR-RBS-2007-00835 and included activities that changed the licensees human performance program such that the human performance error that occurred in September of 2006 is not likely to re-occur. This finding therefore does not have a crosscutting aspect.
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control Rod Inspection Procedure The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving an inadequate control rod inspection procedure. Specifically, the stations procedures only required inspection of a only 20 percent of the control rods that exceeded the inspection criteria, instead of all of them.
The station currently has 18 CR 82M control rods in the reactor core in shutdown locations that have exceeded Westinghouses inspection threshold exposure limits. In response to the inspectors inquries, the licensee reviewed their water chemistry and concluded the current tritium and boron levels indicated there was margin for control rod operability. The licensee intends to monitor the reactor coolant for increasing boron and tritium levels throughout this operating cycle. The licensee placed this issue into their corrective action program as Condition Report CR-RBS-2011-01704.
The finding is more than minor because it is associated with the equipment performance attribute of the reactor safety 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 Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to be of very low safety significance (Green) because it did not result in a loss of system safety function. The inspectors determined that the apparent cause of the performance deficiency was River Bend Stations failure to communicate relevant operating experience to affected internal and external stakeholders. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the operating experience component because the licensee failed to appropriately apply all the CR 82M control rod inspection requirements provided by the control rod
 
vendor [P.2(b)].
Inspection Report# : 2011002 (pdf)
Barrier Integrity Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Standby Gas Treatment Electric Heater Power Output Calculation The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, for an inadequate calculation methodology used in determining standby gas treatment system operability. The inspectors found that the calculation neither considered instrument uncertainty nor applied a proper voltage drop from the breaker to the standby gas treatment system filter train heater. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-07332.
The finding was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone to maintain radiological barrier functionality of standby gas treatment trains, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, operating the standby gas system filter train heaters without sufficient output power is detrimental to the charcoal filters ability to retain radioactive iodine. This could result in a greater amount of radiation release to the environment in the event of an accident. In accordance with Inspection manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the Phase 1 significance determination process screening determined the finding to be only of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the standby gas treatment system. The apparent cause of this finding was the decision to develop an engineering evaluation that did not include instrument uncertainly and did not validate the correct voltage drop between the filter train heater feeder breaker and the heater elements. The finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel failed to use conservative assumptions when developing the modified output power methodology for operation of the standby gas treatment system filter heaters with only 8 of 9 heater elements installed [H.1(b)].
Inspection Report# : 2011004 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Procedure AOP-0027, "Fuel Handling Mishaps" The inspector identified a Green noncited violation of Technical Specification 5.4.1.a, Procedures for River Bend Station fuel handling personnel failing to follow AOP-0027, Fuel Handling Mishaps, when an actual fuel handling event occurred. Instead of entering the AOP, fuel handling personnel continued to move a fuel assembly after equipment damage and potential fuel damage. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-03692.
This failure to follow procedures is a performance deficiency. The performance deficiency is more that minor, and therefore a finding, because it adversely impacted the human performance attribute of the barrier integrity cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, the inspector determined the finding had very low safety significance (Green) because the fuel cladding barrier was potentially degraded but there was no release of radionuclides. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the operating experience component because the licensee failed to implement and institutionalize operating experience through changes to station procedures and training programs [P.2(b)].
Inspection Report# : 2011002 (pdf)
 
Significance:        Mar 31, 2011 Identified By: NRC Item Type: FIN Finding Failure to Follow Fuel Handling Guidelines The inspector identified a finding for failure to follow River Bend Stations Fuel Handling Guideline. A fuel handling event occurred at River Bend Station on January 21, 2011, when a fuel assembly was grappled and raised approximately one foot rather than fully withdrawn from the core. With the fuel assembly only partially withdrawn from the core, the refuel platform was erroneously moved horizontally approximately five feet. This inappropriate stop at one foot followed by inappropriate horizontal movement of the refuel platform with the fuel partially inserted into the core resulted in equipment damage and potential fuel damage. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-03693.
This failure to follow the guideline is a performance deficiency. The performance deficiency is more that minor, and therefore a finding, because it adversely impacted the human performance attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, the inspector determined the finding had very low safety significance (Green) because the fuel cladding barrier was potentially degraded but there was no release of radionuclides. This finding has a crosscutting aspect in the area of human performance associated with the decision making component because the licensee made a safety-significant decision without verifying the validity of underlying assumptions [H.1(b)].
Inspection Report# : 2011002 (pdf)
Emergency Preparedness Significance: SL-IV Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Reduction in ERO Staffing Decreased Emergency Plan Effectiveness The inspector identified a Severity Level IV noncited violation of 10 CFR 50.54(q) for changes to the licensees emergency plans that decreased the effectiveness of those plans without NRC approval. Specifically, the effectiveness of River Bend Station Emergency Plan, Revision 36, was reduced by removal of the Health Physics Communicator position from the emergency response organization. The licensees failure to recognize that Revision 36 decreased the effectiveness of licensee emergency plans was a performance deficiency. The licensee has entered this issue into their corrective action system as CR-RBS-2011-02366.
This finding is more than minor because it has a potential effect on the licensees emergency response capabilities and because the licensee may not be capable of implementing adequate measures to protect the health and safety of the public when the effectiveness of its emergency response organization has been reduced. The finding was evaluated using the NRC Enforcement Policy because it impeded the regulatory process as defined by Manual Chapter 0609, Appendix B, Section 2.2(e). The finding was determined to be Severity Level IV because it decreased the licensees ability to meet or implement a regulatory requirement not related to assessment or notification.
Inspection Report# : 2011002 (pdf)
Occupational Radiation Safety Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Control and Guard the Entrance to a Locked High Radiation Area The inspectors identified a noncited violation of Technical Specification 5.7.2 for failure to properly control and guard
 
a high radiation area with dose rates greater than or equal to 1000 mrem/hr. Specifically, on January 25, 2011, while touring the outside area between the auxiliary building and the radioactive waste building, the inspectors noted that the access gate to a locked high radiation area was open. A guard for the locked high radiation area was positioned in a tent enclosure to the right of the gate, but was not in a position to maintain line-of-sight control of the access to the locked high radiation area. The licensee immediately repositioned the guard and enhanced the tent construction to provide the necessary control for access to the area. The licensee placed this issue into their corrective action program as CR 2011-01154.
The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to properly control access to a high radiation area with dose rates in excess of 1000 mrem/hr had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because it was not associated with ALARA planning or work controls, there was no overexposure, there was no substantial potential for an overexposure, and the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work control, work planning activities, because the individuals failed to consider job site conditions which would impact the ability of the guard to adequately observe the entrance to the locked high radiation area [H.3(a)].
Inspection Report# : 2011002 (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 Significance: N/A May 12, 2011 Identified By: NRC Item Type: FIN Finding River Bend Plant Biennial PI&R Inspection Summary The team determined that the licensees program for identifying, prioritizing, and correcting conditions adverse to quality was effective. With few exceptions, the licensee identified conditions adverse to quality at a low threshold, properly classified and evaluated those conditions, and developed appropriate corrective actions.
The licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items in the corrective action program. The licensee used industry operating experience when performing root cause and apparent cause evaluations. In addition, the licensee performed effective quality assurance audits and self-assessments.
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# : 2011006 (pdf)
Last modified : March 02, 2012
 
River Bend 1 1Q/2012 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Relief Valve Configuration Control Results in a Reactor Downpower The inspectors identified a self-revealing finding for failing to maintain configuration control of the gland seal header relief valves bonnet vent port. The configuration control failure lead to a subsequent decrease in condenser vacuum requiring an unplanned power reduction to maintain adequate condenser vacuum margin. This finding has been entered into the licensees corrective action program as Condition Report CR-RBS-2012-00736.
The failure to maintain configuration control of the glad seal header relief valve was a performance deficiency. The finding was determined to be more than minor because it was associated with the configuration control attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the failure to maintain configuration control resulted in an unplanned down power. Using Inspection Manual Chapter IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems will not be available. The inspectors determined that the apparent cause of this finding was that when the licensee prepared work orders that directed installation of the gland seal header relief valves, they did not comply with procedural requirements to provide plant configuration controls. Therefore, this finding has a cross-cutting aspect in the human performance area associated with the work practice component because the licensee did not define and effectively communicate expectations regarding procedural compliance [H.4 (b)].
Inspection Report# : 2012002 (pdf)
Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Properly Fabricate and Install the mid-Standard Turbine Shaft Brush The inspectors reviewed a self-revealing finding regarding the improper fabrication of a turbine shaft grounding brush that resulted in turbine trip and subsequent reactor scram. The licensee identified the improper fabrication of a turbine shaft grounding brush as the cause of a spurious main turbine over-speed trip signal from an electrical discharge from the turbine shaft. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-9053.
Failure to fabricate the turbine shaft grounding brush in accordance with vendor instructions is a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. Specifically, the improperly fabricated grounding brush resulted in a turbine trip and subsequent reactor scram. The inspectors reviewed the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affecting mitigating equipment, and did not increase the likelihood of a fire or flood. The apparent cause of the performance deficiency was the failure in 2004 to appropriately perform a post maintenance test for the turbine shaft grounding brush modification. Therefore the inspectors did not identify a cross-cutting aspect because the performance deficiency is not reflective of the licensees current performance.
 
Inspection Report# : 2012002 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: FIN Finding Ineffective Corrective Actions on the Main Steam Equalizing Header Drain Bypass Valve Results in an Unplanned Down Power The inspectors identified a self-revealing finding involving inadequate corrective actions in response to a failure in the main steam equalizing header drain bypass valve, resulting in a steam leak and an unplanned plant down power.
Specifically, plant personnel failed to properly address the dual indication on the bypass valve and fluid flow through the valve caused water to flash to steam accelerating pipe wall erosion and piping failure. The licensees immediate corrective actions were to identify, secure, and temporarily repair the steam leak. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-04592.
The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it was not a loss of coolant accident initiator, did not contribute to both the likelihood of an initiating event and the likelihood that mitigating equipment or functions would not be available, nor increase the likelihood of an external event (seismic, flooding, or severe weather event). The apparent cause of the performance deficiency was that the control room and outage control center personnel presumed that the main control room dual indication for the valve was incorrect because previously valve operation successfully closed the valve. Consequently, this finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel did not use a systematic process to assess the condition of the bypass valve, and failed to verify the validity of the underlying assumptions that were used to justify operation with the valve having dual indications [H.1(a)].
Inspection Report# : 2011004 (pdf)
Mitigating Systems Significance:        Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Appropriately Assess and Manage Risk for Internal Flooding Events 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, due to the failure of work control and operations personnel to adequately assess the increase in risk associated with internal flooding events. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2012-00641.
The failure of work control and operations personnel to adequately assess the risk associated with internal flooding is a performance deficiency. The performance deficiency resulted in the overall elevated plant risk placing the plant into the higher licensee-established risk category (Green to Yellow). The performance deficiency 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. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowcharts 1 and 2, the finding was determined to have very low safety significance (Green) because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability deficit was less than 1E-7. The inspectors determined that the apparent cause of the finding was that station personnel routinely failed to review the qualitative risk checklist required by the stations risk management procedure. Therefore, this finding has a cross-cutting aspect
 
in the human performance area associated with the work practice component because the licensee did not define and effectively communicate expectations regarding procedural compliance. [H.4(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Assumptions used in Standby Equipment Room Temperature Analysis The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because, prior to February 7, 2012, the licensee did not verify that assumptions used in confirming that the safety-related battery inverter rooms would remain below their design basis temperature limits during a design basis event agreed with the as-built condition of the plant. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2012-01046.
The inspectors determined that the failure to verify that design documents match the actual configuration of the plant is a performance deficiency. The finding was more than minor because it adversely affects the Mitigating Systems Cornerstone objective of equipment performance to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee had not verified assumptions that ensure the standby switchgear room air conditioning system would reliably maintain the standby equipment rooms below the design temperature limits. Using Inspection Manual Chapter 0609, Attachment 4, "Initial Screening and Characterization of Findings," the finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function, nor actual loss of safety function of a single train, and it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that this issue has a cross-cutting aspect in the area of human performance decision-making regarding nonconservative assumptions. When the licensee conducted the flow balance test, they assumed that measuring air inflow alone was sufficient, but did not check that the doors gaps were allowing a sufficient amount of warm air to exit standby equipment rooms and be circulated back to the general areas [H.1(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Appropriately Set Reactor Core Isolation Cooling Flow Controller High Output Limit The inspectors identified a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Procedures, when the reactor core isolation cooling turbine tripped on mechanical over speed. Troubleshooting determined the cause was an improperly tuned flow controller. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2012-01188 and CR-RBS-2012-01262.
The failure to provide specific flow controller tuning instructions for the reactor core isolation cooling turbine flow controller was a performance deficiency. The finding was more than minor in accordance with Appendix B, "Issue Screening," of Inspection Manual Chapter IMC 0612, "Power Reactor Inspection Reports," because the finding was associated with the equipment performance 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, improper tuning of the reactor core isolation cooling controller impacted operability and availability of the reactor core isolation cooling system. The inspectors performed a Phase 1 significance determination process review of this finding per Inspection Manual Chapter IMC 0609, Attachment 4, "Initial Screening and Characterization of Findings." In accordance with Table 4a, "Characterization Worksheet for IE, MS, and BI Cornerstones," the finding represented a loss of system safety function. Therefore, a Region IV senior reactor analyst used Inspection Manual Chapter IMC 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," to review the finding using the Standardized Plant Analysis Risk (SPAR) model for River Bend Station. The Phase 3 analysis determined the Delta-CDF was 4.68E-7/yr. For a 7-month exposure, the incremental conditional core damage probability is 2.73E-7. The majority of the risk came from sequences involving a loss of feedwater (48 percent) and a loss of offsite power (33 percent). Consequently, the analyst determined that the risk associated with the performance deficiency was very low (green). The inspectors determined the apparent cause of this finding was the failure to perform a post maintenance test to identify that the
 
high output limit was not properly set by the maintenance work instruction. Therefore, this finding has cross-cutting aspect in the area of human performance associated with the resources component due to less than adequate work package testing instruction. [H.2(c)].
Inspection Report# : 2012002 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Standby Service Water Pump Motor Lubrication Deficiencies The inspectors identified a Green non-cited violation of Technical Specification 5.4.1.a, because the station did not establish appropriate maintenance procedures to lubricate standby service water pump lower motor bearings.
Specifically, the inspectors found a legacy of improper maintenance practices involving lubrication of the standby service water pump motor lower bearings going back to 1986. This included mixing of incompatible greases without change evaluations, lubrication techniques that did not comply with pump motor vendor manual or EPRI guidance, improper volume of greases added to the bearings, and improper preventive maintenance frequency for performing re-greasing of the bearings. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-08367.
This performance deficiency is more-than-minor and is therefore a finding because if left uncorrected, this performance deficiency has the potential to lead to a more significant safety concern. Specifically, if the subject work orders are not corrected, future work activities that grease the subject bearings in accordance with those work orders may not grease the bearings adequately, which may result in common-cause failures of the station service water pumps. Because this finding was identified while the unit was operating, the inspectors used MC 0609 Appendix A to assess its risk significance. In accordance with that Appendix, the finding screened as green (of very low safety significance) because it was not a design or qualification deficiency; it did not represent a loss of system safety function; and it did not screen as potentially risk-significant due to seismic, flooding, or severe weather initiating events. The inspectors determined that the apparent cause of this finding was failure to include the appropriate scope of information in the work instructions due to overconfidence and lack of adequate review by engineering staff.
Specifically, the system engineer who developed the revised instructions failed to develop appropriate steps with adequate detail to appropriately perform the task and the field engineer failed to stop work and discuss the issue with the system engineer that developed the work instructions. Therefore, the finding has a crosscutting aspect in the area of human performance associated with work practices, because engineering personnel failed to use the applicable human error prevention techniques [H.4(a)].
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Control Building Chiller System The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of the control building chilled water system. Specifically, the inspectors determined that the station had failed to track system unavailability following the systems classification of a high risk system and did not monitor the system at the train level, ultimately masking the performance of individual trains. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-07332.
The finding was more than minor since violations of 10 CFR 50.65(a)(2) necessarily involve degraded system performance which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the finding did not lead to an actual loss of safety function of the system or cause a component to be inoperable, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the cause of the finding was the lack of management oversight.
Following the issuance of River Bend Station Probabilistic Risk Assessment interim Revision 4a, several personnel functioned as the maintenance rule coordinator and control building chilled water system engineer. During this period, station management did not ensure sufficient knowledge transfer for effective maintenance rule implementation.
Therefore, this finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to ensure supervisory and management oversight of work activities such that
 
nuclear safety is supported [H.4(c)].
Inspection Report# : 2011005 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Testing of Division I and Division III Standby Diesel Generators The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, prior to October 27, 2011, the licensee failed to ensure surveillance testing procedures of Division I and III standby diesel generators incorporated the correct acceptance limits for maximum expected load at max frequency and voltage specified in design basis documents. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07132, CR-RBS-2011-07294, and CR-RBS-2011-07518.
The team determined that the failure to ensure that the test procedures required to demonstrate that Division I and Division III standby diesel generators will perform satisfactorily in service incorporated the requirements and acceptance limits contained in applicable design documents was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee could not ensure that the standby diesel generators would reliably provide power for the maximum expected post-accident loads including maximum frequency and voltage. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1
- Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of condition
[P.1(c)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Conservative Design Assumptions in the Ultimate Heat Sink Inventory Calculation The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 27, 2011, the licensee failed to assure that the design basis information for expected heat loads to the ultimate heat sink was correctly translated into the ultimate heat sink 30-day inventory analysis. The analysis used a less conservative, frictionless form of the conservation of energy equation to determine heat load in the standby service water system during a 30-day design basis event. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07430 and CR-RBS-2011-07654.
The team determined that the failure to correctly translate expected heat loads into the ultimate heat sink inventory analysis was a performance deficiency. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to undesired consequences. Specifically, the neglect of friction heat load in the ultimate heat sink analysis system resulted in a condition where there was reasonable doubt on the operability of a system to meet its 30-day mission time without a makeup water source. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was
 
a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Specifically, the licensees revised analysis to determine operability removed overly conservative assumptions for operating the low pressure core spray pump for 30 days to account for the friction heat load added to the system. The finding has a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to thoroughly evaluate problems such that the resolutions address cause and extent of condition [P.1(c)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Residual Heat Removal Heat Exchanger Testing Frequency The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, from October 1998 to October 27, 2011, the licensee failed to establish a NRC Generic Letter 89-13 test program which incorporated a final test frequency for the residual heat removal heat exchangers and perform an adequate trending analysis upon which to base a final test frequency. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07713.
The team determined that failure to establish a NRC Generic Letter 89-13 test program which incorporated a final testing frequency of the residual heat removal heat exchangers was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inappropriate test frequency affected the licensees ability to ensure residual heat removal heat exchangers, when called upon, were available and capable to reliably perform as expected. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determine to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding did not have a crosscutting aspect because the most significance contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Monitoring Standby Service Water System Leakage The team identified a Green, noncited violation of 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, which states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
Specifically, prior to October 27, 2011, the licensee failed to provide appropriate quantitative or qualitative acceptance criteria in station and abnormal operating procedures to determine if actions for leak detection were satisfactorily accomplished to protect the standby service water system and ultimate heat sink during design basis events. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07555.
The team determined that the failure to include appropriate acceptance criteria for leak detection in abnormal operating procedures for the standby service water system and ultimate heat sink was a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inadequate
 
procedure guidance could lead to operators not recognizing conditions that would degrade the availability of the standby service water system. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 -
Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance: SL-IV Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for Change to Ultimate Heat Sink Inventory Requirements The team identified a Severity Level IV, noncited violation of 10 CFR 50.59, Changes, Tests and Experiments which states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if this activity would; result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the final safety analysis report (as updated). Specifically, from December 16, 2002, to October 27, 2011, the licensee changed the design basis of the ultimate heat sink inventory requirements to provide a 30-day cooling water supply without makeup capability to providing a less than 30-day cooling water supply with makeup capability without obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR 2011-07674.
The team determined that the failure to obtain a license amendment prior to implementing a proposed change, test or experiment to the ultimate heat sink requirements was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding has the ability to impact the regulatory process. The finding was more than minor because it involved a change to the updated final safety analysis report description where there was a reasonable likelihood that the change would require NRC approval. In accordance with the NRC Enforcement Policy, the team used insights from MC 0609, Significance Determination Process, to determine the final significance of the finding. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 -
Initial Screening and Characterization of Findings, the finding represented a loss of system safety function in that the ultimate heat sink could not meet its 30-day mission time to provide decay heat removal. Therefore, a Phase 2 evaluation was necessary. The significance of the finding could not be assessed quantitatively through a Phase 2 or Phase 3 analysis. Consequently, an assessment was performed in accordance with IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The finding was determined to have very low safety significance because the frequency of events that would require long term use of the ultimate heat sink is very low and the difference in the failure probability to replenish the ultimate heat sink in 10 days versus 30 days is very small. This was because an early depletion of the inventory would be easily detected and would become a priority. At the time that replenishment would be needed, plant conditions should be stable and local transportation arteries should be restored. Therefore, since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance:      Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Abnormal Procedure for Reducing Loads on Standby Diesel Generators The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, and drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, prior to October 27, 2011, the licensee failed to include appropriate qualitative and quantitative acceptance criteria in abnormal operating procedures for control room operators to recognize the need to reduce loads on the standby diesel generators during design basis accidents. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07716.
 
The team determined that the failure to include appropriate quantitative or qualitative acceptance criteria in abnormal operating procedures for control room operators to recognize the need to reduce loads on the standby diesel generators during design basis accidents was a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, a control room operating crews failure to recognize the need to reduce loads to prevent the standby diesel generator failure during design basis accidents adversely affected the reliability of the standby diesel generators. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety for the correct training of licensed operator personnel [H.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency and Abnormal Procedures for Standby Diesel Generator Fail to Load Sequences The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, and drawings shall include appropriate qualitative and quantitative criteria for determining that important activities have been satisfactorily accomplished.
Specifically, prior to October 27, 2011, the licensee failed to include appropriate quantitative or qualitative acceptance criteria in procedures for control room operators to recognize and recover a standby diesel generator that starts but fails to load with the remaining standby diesel generator out of service during a loss-of-offsite-power event. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07716, CR-RBS-2011-07717, and CR-RBS-2011-07718.
The team determined that the failure to include appropriate quantitative or qualitative acceptance criteria to determine that important activities are satisfactorily accomplished in emergency and abnormal operating procedures used during loss-of-offsite-power events was a performance deficiency. The finding was more than minor because it is associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, a control room operator crews failure to diagnose recoverable conditions adversely affected the availability of standby diesel generators during a loss-of-offsite-power event. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of problem identification and resolution, operating experience component, because the licensee did not implement and institutionalize operating experience through changes to station processes, procedures, equipment, and training programs [P.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Simulator Fidelity for Emergency Diesel Generator Loading The team identified a Green, noncited violation of 10 CFR 55.46(c)(1), Simulation Facilities, which states, in part,
 
that a plant-referenced simulator used for the administration of the operating test must demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond. Specifically, prior to October 27, 2011, the River Bend Station simulator did not demonstrate the expected plant response for standby diesel generator loading during accident conditions to which the simulator was designed to respond. The electrical loading on the emergency diesel generator in the simulator was approximately 800 kW less than the expected full load for the diesel generator. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07682.
The team determined that the failure of the plant-referenced simulator to demonstrate expected plant response for standby diesel generator loading during accident conditions to which the simulator has been designed to respond was a performance deficiency. The finding was more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the incorrect simulator response adversely affected the control room operator crews capability to assess standby diesel generator loading conditions. In accordance with NRC Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets and the associated Appendix I, the finding was determined to be of very low safety significance (Green). Specifically, Manual Chapter 0609, Appendix I, Operator Requalification Human Performance Significance Determination Process, block 12, establishes a Green finding for failure to correctly replicate the plants response on the simulator that either has the potential to cause or actually causes negative training to operators. Negative training did occur for this finding because operators thought they had electrical load margin on the emergency diesel generators when the diesels were actually fully loaded with minimal margin without securing other equipment. This finding had a crosscutting aspect in the area of human performance, resources component, in that the licensee did not ensure that equipment (plant-referenced simulator) was adequate to assure nuclear safety for the correct training of licensed operator personnel [H.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Failure to Track and Document Plant Equipment Oil Usage The inspectors identified a finding for the failure to properly document equipment oil additions in the oil lubrication accountability log per General Maintenance Procedure GMP-0015, Lubrication Procedure. To correct the programmatic deficiencies, the station revised General Maintenance Procedure GMP-0015 instructions to enhance and amplify the requirement to record all oil additions in the lubrication accountability log, revise preventative maintenance tasks that sample or change oil to explicitly state record oil additions in the lubrication accountability log, and to brief station personnel concerning changes to General Maintenance Procedure GMP-0015. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-02883.
The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency; did not represent either a loss of system safety function, an actual loss of safety function of a single train, or an actual loss of safety function; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the apparent cause of the performance deficiency was incomplete work package instructions that did not explicitly state to record oil additions in the lubrication accountability log per General Maintenance Procedure GMP-0015, thereby making equipment operability conclusions based on incomplete monitored trends suspect and potentially inaccurate. Consequently, this finding has a cross-cutting aspect in the human performance area associated with the resources component because the stations work packages lacked the necessary instructions to adequately control the lubrication monitoring program
[H.2(c)].
Inspection Report# : 2011003 (pdf)
Significance:        May 12, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation
 
Failure to Take Corrective Action for Service-Induced Failures of Gould J-series Relays The inspectors reviewed a self-revealing green noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take corrective action to address service-induced failures of Gould J series relays. In response, the licensee initiated condition report CR RBS 2010 06032 to ensure that appropriate levels of preventive maintenance are performed on high-critical components.
The performance deficiency was the licensee's failure to take adequate corrective actions to address service-induced failures of the high-critical, high-duty-cycle Gould J series relay designated as EHS MCC16B6D 33X1. This performance deficiency was determined to be more than minor and was therefore a finding because it impacted the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding had very low safety significance because the finding was not a design or qualification deficiency confirmed not to result in a loss of operability, did not represent a loss of system safety function, did not represent 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. Because the apparent cause of this finding was the licensees misclassification of the failed relay within the preventive maintenance optimization program in 2008, and because the licensees performance in that program was not reflective of current licensee performance, no cross-cutting aspect was assigned to this finding.
Inspection Report# : 2011006 (pdf)
Barrier Integrity Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Maintenance Instructions used for Suppression Pool Cooling Isolation Valve Maintenance The inspectors identified a Green, self-revealing non-cited violation of Technical Specification 5.4.1.a, Procedures, for inadequate maintenance procedures to properly assemble containment isolation valves on the suppression pool cooling system. This resulted in a failure of the suppression pool cooling systems outboard containment isolation valve marriage coupling that ensures the valve stem is connected to the valve actuator. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2011-09171.
The failure to establish adequate work instructions to assemble the suppression pool cleanup system isolation valves is a performance deficiency. The inspectors determined that the finding was more than minor because it is associated with the Barrier Integrity Cornerstone attribute of Systems, Structures, and Components and Barrier Performance, and affected the cornerstone objective of providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations. Using the Phase 1 SDP worksheet for the barrier integrity cornerstone, the inspectors answered no to all four screening questions under the containment barrier column. Specifically, the affected penetration did not represent an actual open pathway in the physical integrity of reactor containment due to an operable and functionally redundant containment isolation valve in the suppression pool cooling piping penetration. The apparent cause of the finding was the failure of the planning department to recognize and develop design documentation to identify the set screw size and starting material necessary to determine the appropriate set screw torque for work affecting safety related equipment. The inspectors determined the finding had a cross cutting aspect in the human performance, area associated with the resources component because of the lack of complete accurate and up to date design documentation associated with the work package development. [H.2(c)].
Inspection Report# : 2012002 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Standby Gas Treatment Electric Heater Power Output Calculation
 
The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, for an inadequate calculation methodology used in determining standby gas treatment system operability. The inspectors found that the calculation neither considered instrument uncertainty nor applied a proper voltage drop from the breaker to the standby gas treatment system filter train heater. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-07332.
The finding was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone to maintain radiological barrier functionality of standby gas treatment trains, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, operating the standby gas system filter train heaters without sufficient output power is detrimental to the charcoal filters ability to retain radioactive iodine. This could result in a greater amount of radiation release to the environment in the event of an accident. In accordance with Inspection manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the Phase 1 significance determination process screening determined the finding to be only of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the standby gas treatment system. The apparent cause of this finding was the decision to develop an engineering evaluation that did not include instrument uncertainly and did not validate the correct voltage drop between the filter train heater feeder breaker and the heater elements. The finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel failed to use conservative assumptions when developing the modified output power methodology for operation of the standby gas treatment system filter heaters with only 8 of 9 heater elements installed [H.1(b)].
Inspection Report# : 2011004 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform a Radiation Survey Inspectors reviewed a self-revealing non-cited violation of 10 CFR 20.1501(a) for the failure to perform a radiation survey. A survey was not completed after two contaminated valves were transferred from the 98-foot elevation of the main steam tunnel to the radwaste area. During shift turnovers, workers responsible for transferring the valves did not understand that they needed to remove two buckets, and perform a survey after completing the valve transfer.
Consequently, a bucket with highly contaminated water and residual was left in the tunnel causing radiation levels as high as 300 millirem per hour. This resulted in an unposted high radiation area. The licensee entered the issue into the corrective actions program as Condition Report CR-RBS-2011-01552.
The failure to perform a radiation survey to evaluate the radiological conditions is a performance deficiency. The finding is more than minor because it negatively impacted the Occupational Radiation Safety cornerstones attribute of program and process, in that the lack of a post-work survey did not ensure exposure control for workers. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of 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 the ability to assess dose was not compromised. The finding has a Human Performance cross-cutting component associated with the aspect of work practices because expectations regarding procedural compliance for post-job radiation surveys were ineffective [H.4(b)].
Inspection Report# : 2012002 (pdf)
 
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Control Access to a High Radiation Area Inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.7.1(c), resulting from the licensees failure to control access to a high radiation area. Specifically, a carpenter entered a high radiation area in the main steam tunnel near valve V112 without proper authorization before a health physics technician completed radiation surveys and received an unexpected alarming dosimeter reading of 110 millirem per hour. The carpenter had not been briefed that dose rates in the area measured 140 millirem per hour. He had been instructed not to perform any work before the health physics technician surveyed the area, but River Bend did not make it clear enough that he was to follow all health physics instructions. The licensee entered the issue into the corrective actions program as Condition Report CR-RBS-2011-01426 and the worker was counseled.
The failure to control access to a high radiation area was a performance deficiency. The finding was more than minor because it was associated with the occupational radiation safety attribute of exposure control and affected the cornerstone objective in that not controlling a high radiation area could increase personal exposure. In addition, this type of issue is addressed in Example 6.h of IMC 0612, Appendix E, Examples of Minor Issues. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The finding has a Human Performance cross-cutting component associated with the aspect of work practices because expectations regarding supervisory and management oversight of work activities, including contractors to ensure that safety is supported were not met [H.4(c)].
Inspection Report# : 2012002 (pdf)
Public Radiation Safety Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Digital Radiation Monitoring System The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of the digital radiation monitoring system. Specifically, the maintenance rule expert panel performed an inadequate analysis after the digital radiation monitoring system exceeded the condition monitoring criteria by failing to follow the procedural requirements of EN-DC-206 to have cause evaluations for system failures so that maintenance preventability could be properly evaluated. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2011-00485.
The inspectors determined that the failure to adequately monitor the performance of the digital radiation monitoring system is a performance deficiency. The inspectors reviewed Inspection Manual Chapter (IMC) 0612 and determined that the finding is more than minor because the finding is associated with the plant facilities/equipment and instrumentation attribute (reliability of process radiation monitors) of the radiation safety cornerstone (public radiation safety) and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian use. The finding was assessed using the IMC 0609, Appendix D, Public Radiation SDP, and because there was no failure to implement the effluent program, the finding was determined to be of very low safety significance (Green). The inspectors reviewed the apparent cause of this finding and found that the oversight of the maintenance rule program was adversely affected by personnel changes and lack of effective turnover. Therefore, the finding has a cross-cutting aspect in the human performance area and resources component because the licensee failed to ensure that maintenance rule program personnel were trained and sufficiently qualified to perform their duties in an effective manner [H.2(b)].
Inspection Report# : 2012002 (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 May 12, 2011 Identified By: NRC Item Type: FIN Finding River Bend Plant, 2011, Biennial PI&R Inspection Summary The team determined that the licensees program for identifying, prioritizing, and correcting conditions adverse to quality was effective. With few exceptions, the licensee identified conditions adverse to quality at a low threshold, properly classified and evaluated those conditions, and developed appropriate corrective actions.
The licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items in the corrective action program. The licensee used industry operating experience when performing root cause and apparent cause evaluations. In addition, the licensee performed effective quality assurance audits and self-assessments.
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# : 2011006 (pdf)
Last modified : May 29, 2012
 
River Bend 1 2Q/2012 Plant Inspection Findings Initiating Events Significance:        Jun 29, 2012 Identified By: NRC Item Type: FIN Finding Failure to Follow Procedure to Protect Sensitive Plant Areas The inspectors identified a finding for failure to follow Operating System Procedure OSP-0048, "Switchyard, Transformer Yard, and Sensitive Equipment Controls." Specifically, the licensee failed to appropriately consider the plant impact when planning and approving work in the main transformer yard and switchyard potentially introducing unacceptable risk to plant operations contrary to OSP-0048 administrative controls. This issue was entered into the licensees corrective action program as Condition Reports CR-RBS-2012-02479, CR-RBS-2012-02821, and CR-RBS-2012-04129.
The finding was more than minor in accordance with Appendix B, "Issue Screening," of Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," because the finding was associated with the protection against external events attribute of the Initiating Events cornerstone and affected 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. Specifically, the routine failure to integrate switchyard and transformer yard work into the River Bend work process increased the likelihood that unintended, uncoordinated maintenance and test activities could reduce the diversity of electrical power and cause inadvertent reductions in nuclear plant defense-in-depth. The inspectors performed a Phase 1 significance determination process review of this finding per Inspection Manual Chapter 0609, , "Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) since the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, nor did it contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, and the finding did not increase the likelihood of a fire or internal or external flooding. The inspectors determined the apparent cause of this finding was a lack of management oversight of station work activities. Therefore, this finding has a cross-cutting aspect in the area of human performance associated with the work practices component because station management failed to provide proper oversight of the process to protect sensitive areas of the plant [H.4(c)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: FIN Finding Failure to Implement Severe Weather Operations Procedure The inspectors identified a finding that involved failure to implement a procedure to protect the plant during adverse weather conditions. Specifically, appropriate equipment walkdowns and corrective actions were not performed to protect equipment important to safety from severe weather risks in a timely manner. The concerns were documented in Condition Report CR-RBS-2012-02387.
The finding was determined to be of very low safety significance (Green) since the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, nor did it contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, and the finding did not increase the likelihood of a fire or internal or external flooding. The inspectors determined the apparent cause of this finding was operations expectation that excellent housekeeping nominally exists in the switchyard and transformer yard. Therefore, there was no need to dispatch personnel to verify housekeeping because that action would risk personnel safety. The status of an unsecured ladder in the transformer yard is evidence that up to date information is essential to confirm whether housekeeping is satisfactory. Therefore, the finding has a cross-
 
cutting aspect in the area of human performance associated with the decision-making component because the station did not demonstrate that nuclear safety was an overriding priority because it failed to implement the roles and authorities in their severe weather operations procedure [H.1(a)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Properly Assemble Turbine Control Valve Push Rod-Spring Housing Coupling The inspectors reviewed a self-revealing finding associated with main turbine control valve number 3 unexpectedly closing. In response, operators reduced reactor power to 90 percent. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-02773.
The finding was more than minor because it was associated with the Initiating Events cornerstone attribute of design control and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability by resulting in a plant downpower and subsequent planned outage for repair activities. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affecting mitigating equipment, and did not increase the likelihood of a fire or flood. The inspectors did not identify a cross cutting aspect because the performance deficiency is not indicative of the licensees current performance.
Inspection Report# : 2012003 (pdf)
Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Relief Valve Configuration Control Results in a Reactor Downpower The inspectors identified a self-revealing finding for failing to maintain configuration control of the gland seal header relief valves bonnet vent port. The configuration control failure lead to a subsequent decrease in condenser vacuum requiring an unplanned power reduction to maintain adequate condenser vacuum margin. This finding has been entered into the licensees corrective action program as Condition Report CR-RBS-2012-00736.
The failure to maintain configuration control of the glad seal header relief valve was a performance deficiency. The finding was determined to be more than minor because it was associated with the configuration control attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the failure to maintain configuration control resulted in an unplanned down power. Using Inspection Manual Chapter IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems will not be available. The inspectors determined that the apparent cause of this finding was that when the licensee prepared work orders that directed installation of the gland seal header relief valves, they did not comply with procedural requirements to provide plant configuration controls. Therefore, this finding has a cross-cutting aspect in the human performance area associated with the work practice component because the licensee did not define and effectively communicate expectations regarding procedural compliance [H.4 (b)].
Inspection Report# : 2012002 (pdf)
Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Properly Fabricate and Install the mid-Standard Turbine Shaft Brush The inspectors reviewed a self-revealing finding regarding the improper fabrication of a turbine shaft grounding brush
 
that resulted in turbine trip and subsequent reactor scram. The licensee identified the improper fabrication of a turbine shaft grounding brush as the cause of a spurious main turbine over-speed trip signal from an electrical discharge from the turbine shaft. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-9053.
Failure to fabricate the turbine shaft grounding brush in accordance with vendor instructions is a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. Specifically, the improperly fabricated grounding brush resulted in a turbine trip and subsequent reactor scram. The inspectors reviewed the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affecting mitigating equipment, and did not increase the likelihood of a fire or flood. The apparent cause of the performance deficiency was the failure in 2004 to appropriately perform a post maintenance test for the turbine shaft grounding brush modification. Therefore the inspectors did not identify a cross-cutting aspect because the performance deficiency is not reflective of the licensees current performance.
Inspection Report# : 2012002 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: FIN Finding Ineffective Corrective Actions on the Main Steam Equalizing Header Drain Bypass Valve Results in an Unplanned Down Power The inspectors identified a self-revealing finding involving inadequate corrective actions in response to a failure in the main steam equalizing header drain bypass valve, resulting in a steam leak and an unplanned plant down power.
Specifically, plant personnel failed to properly address the dual indication on the bypass valve and fluid flow through the valve caused water to flash to steam accelerating pipe wall erosion and piping failure. The licensees immediate corrective actions were to identify, secure, and temporarily repair the steam leak. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-04592.
The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it was not a loss of coolant accident initiator, did not contribute to both the likelihood of an initiating event and the likelihood that mitigating equipment or functions would not be available, nor increase the likelihood of an external event (seismic, flooding, or severe weather event). The apparent cause of the performance deficiency was that the control room and outage control center personnel presumed that the main control room dual indication for the valve was incorrect because previously valve operation successfully closed the valve. Consequently, this finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel did not use a systematic process to assess the condition of the bypass valve, and failed to verify the validity of the underlying assumptions that were used to justify operation with the valve having dual indications [H.1(a)].
Inspection Report# : 2011004 (pdf)
Mitigating Systems Significance:        Jun 29, 2012
 
Identified By: NRC Item Type: NCV NonCited Violation High Pressure Core Spray Diesel Generator Bearing Lubrication Deficiencies The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failing to correct a condition adverse to quality for lubricating the high pressure core spray diesel generator bearings.
The station documented the finding in Condition Report CR-RBS-2012-02666.
This performance deficiency was more than minor and was a finding because, if left uncorrected, inadequate lubrication work instruction could cause bearing failure due to inadequate lubrication or generator winding failure due to grease intrusion into the electrical windings in the generator. The significance of this finding was evaluated using a Phase 1 significance determination process screening and was determined to be of 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 seismic, flooding, or severe weather initiating events. The apparent reason the initial condition report was closed without correcting the work instruction to lubricate the high pressure core spray diesel generator bearings was that personnel who prepared and approved the operability evaluation were focused on proving operability not correcting a condition adverse to quality. Their focus was specific to the components ability to perform its function and not on completely identifying the issue in the corrective action program. Therefore, the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the station did not identify this issue completely, accurately, and in a timely manner commensurate with its safety significance [P.1(a)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Specify Manual Actions for Safety Relief Valve Operations During a Station Blackout Event The inspectors identified a non-cited violation of 10 CFR 50.63, Loss of All Alternating Current, paragraph (a) (2),
which states, in part, The reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. The capability for coping with a station blackout of specified duration shall be determined by an appropriate coping analysis. Licensees are expected to have the baseline assumptions, analyses, and related information used in their coping evaluations available for NRC review. Specifically, from November 1985 to May 17, 2012, the licensee failed to specify actions while ac power is unavailable to ensure that safety relief valves provided sufficient capacity and capability to ensure appropriate containment integrity is maintained during a station blackout event. This violation has been entered into the corrective action program as Condition Report CR-RBS-2012-03376.
The inspectors determined that failure to specify actions for safety relief valve operation in procedures in accordance with NUMARC-8700 was a performance deficiency. The finding was more than minor because it adversely affected the procedure quality attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to respond to undesirable consequences. Specifically, the station blackout coping procedures did not specify actions that would ensure the heat capacity temperature limit for the suppression pool would not be exceeded during the station blackout coping period. Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the inspectors determined that the Mitigating Systems Cornerstone was affected because the finding could cause degradation of core decay heat removal. Using Table 4a from the Phase 1 worksheet, the inspectors determined that the finding represents a loss of safety function; therefore, a Phase 2 analysis was necessary. However, the inspectors determined that a Phase 2 analysis was not sufficient to assess significance because of the complexity of the finding. Therefore, a Phase 3 analysis was necessary. The result of the Phase 3 analysis determined that the change in core-damage-frequency (?CDF) for the performance deficiency was 2.4E-7 or very low safety significance (Green). The senior reactor analyst determined that the change in large-early-release-frequency (?LERF) was 4.8E-8 or very low safety significance (Green). No cross-cutting aspect was identified because the most significant contributor was not indicative of current licensee performance (Section 4OA5).
Inspection Report# : 2012003 (pdf)
 
Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Appropriately Assess and Manage Risk for Internal Flooding Events 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, due to the failure of work control and operations personnel to adequately assess the increase in risk associated with internal flooding events. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2012-00641.
The failure of work control and operations personnel to adequately assess the risk associated with internal flooding is a performance deficiency. The performance deficiency resulted in the overall elevated plant risk placing the plant into the higher licensee-established risk category (Green to Yellow). The performance deficiency 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. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowcharts 1 and 2, the finding was determined to have very low safety significance (Green) because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability deficit was less than 1E-7. The inspectors determined that the apparent cause of the finding was that station personnel routinely failed to review the qualitative risk checklist required by the stations risk management procedure. Therefore, this finding has a cross-cutting aspect in the human performance area associated with the work practice component because the licensee did not define and effectively communicate expectations regarding procedural compliance. [H.4(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Assumptions used in Standby Equipment Room Temperature Analysis The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because, prior to February 7, 2012, the licensee did not verify that assumptions used in confirming that the safety-related battery inverter rooms would remain below their design basis temperature limits during a design basis event agreed with the as-built condition of the plant. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2012-01046.
The inspectors determined that the failure to verify that design documents match the actual configuration of the plant is a performance deficiency. The finding was more than minor because it adversely affects the Mitigating Systems Cornerstone objective of equipment performance to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee had not verified assumptions that ensure the standby switchgear room air conditioning system would reliably maintain the standby equipment rooms below the design temperature limits. Using Inspection Manual Chapter 0609, Attachment 4, "Initial Screening and Characterization of Findings," the finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function, nor actual loss of safety function of a single train, and it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that this issue has a cross-cutting aspect in the area of human performance decision-making regarding nonconservative assumptions. When the licensee conducted the flow balance test, they assumed that measuring air inflow alone was sufficient, but did not check that the doors gaps were allowing a sufficient amount of warm air to exit standby equipment rooms and be circulated back to the general areas [H.1(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Appropriately Set Reactor Core Isolation Cooling Flow Controller High Output Limit
 
The inspectors identified a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Procedures, when the reactor core isolation cooling turbine tripped on mechanical over speed. Troubleshooting determined the cause was an improperly tuned flow controller. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2012-01188 and CR-RBS-2012-01262.
The failure to provide specific flow controller tuning instructions for the reactor core isolation cooling turbine flow controller was a performance deficiency. The finding was more than minor in accordance with Appendix B, "Issue Screening," of Inspection Manual Chapter IMC 0612, "Power Reactor Inspection Reports," because the finding was associated with the equipment performance 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, improper tuning of the reactor core isolation cooling controller impacted operability and availability of the reactor core isolation cooling system. The inspectors performed a Phase 1 significance determination process review of this finding per Inspection Manual Chapter IMC 0609, Attachment 4, "Initial Screening and Characterization of Findings." In accordance with Table 4a, "Characterization Worksheet for IE, MS, and BI Cornerstones," the finding represented a loss of system safety function. Therefore, a Region IV senior reactor analyst used Inspection Manual Chapter IMC 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," to review the finding using the Standardized Plant Analysis Risk (SPAR) model for River Bend Station. The Phase 3 analysis determined the Delta-CDF was 4.68E-7/yr. For a 7-month exposure, the incremental conditional core damage probability is 2.73E-7. The majority of the risk came from sequences involving a loss of feedwater (48 percent) and a loss of offsite power (33 percent). Consequently, the analyst determined that the risk associated with the performance deficiency was very low (green). The inspectors determined the apparent cause of this finding was the failure to perform a post maintenance test to identify that the high output limit was not properly set by the maintenance work instruction. Therefore, this finding has cross-cutting aspect in the area of human performance associated with the resources component due to less than adequate work package testing instruction. [H.2(c)].
Inspection Report# : 2012002 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Standby Service Water Pump Motor Lubrication Deficiencies The inspectors identified a Green non-cited violation of Technical Specification 5.4.1.a, because the station did not establish appropriate maintenance procedures to lubricate standby service water pump lower motor bearings.
Specifically, the inspectors found a legacy of improper maintenance practices involving lubrication of the standby service water pump motor lower bearings going back to 1986. This included mixing of incompatible greases without change evaluations, lubrication techniques that did not comply with pump motor vendor manual or EPRI guidance, improper volume of greases added to the bearings, and improper preventive maintenance frequency for performing re-greasing of the bearings. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-08367.
This performance deficiency is more-than-minor and is therefore a finding because if left uncorrected, this performance deficiency has the potential to lead to a more significant safety concern. Specifically, if the subject work orders are not corrected, future work activities that grease the subject bearings in accordance with those work orders may not grease the bearings adequately, which may result in common-cause failures of the station service water pumps. Because this finding was identified while the unit was operating, the inspectors used MC 0609 Appendix A to assess its risk significance. In accordance with that Appendix, the finding screened as green (of very low safety significance) because it was not a design or qualification deficiency; it did not represent a loss of system safety function; and it did not screen as potentially risk-significant due to seismic, flooding, or severe weather initiating events. The inspectors determined that the apparent cause of this finding was failure to include the appropriate scope of information in the work instructions due to overconfidence and lack of adequate review by engineering staff.
Specifically, the system engineer who developed the revised instructions failed to develop appropriate steps with adequate detail to appropriately perform the task and the field engineer failed to stop work and discuss the issue with the system engineer that developed the work instructions. Therefore, the finding has a crosscutting aspect in the area of human performance associated with work practices, because engineering personnel failed to use the applicable human error prevention techniques [H.4(a)].
 
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Control Building Chiller System The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of the control building chilled water system. Specifically, the inspectors determined that the station had failed to track system unavailability following the systems classification of a high risk system and did not monitor the system at the train level, ultimately masking the performance of individual trains. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-07332.
The finding was more than minor since violations of 10 CFR 50.65(a)(2) necessarily involve degraded system performance which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the finding did not lead to an actual loss of safety function of the system or cause a component to be inoperable, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the cause of the finding was the lack of management oversight.
Following the issuance of River Bend Station Probabilistic Risk Assessment interim Revision 4a, several personnel functioned as the maintenance rule coordinator and control building chilled water system engineer. During this period, station management did not ensure sufficient knowledge transfer for effective maintenance rule implementation.
Therefore, this finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to ensure supervisory and management oversight of work activities such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2011005 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Testing of Division I and Division III Standby Diesel Generators The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, prior to October 27, 2011, the licensee failed to ensure surveillance testing procedures of Division I and III standby diesel generators incorporated the correct acceptance limits for maximum expected load at max frequency and voltage specified in design basis documents. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07132, CR-RBS-2011-07294, and CR-RBS-2011-07518.
The team determined that the failure to ensure that the test procedures required to demonstrate that Division I and Division III standby diesel generators will perform satisfactorily in service incorporated the requirements and acceptance limits contained in applicable design documents was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee could not ensure that the standby diesel generators would reliably provide power for the maximum expected post-accident loads including maximum frequency and voltage. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1
- Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of condition
[P.1(c)].
 
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Conservative Design Assumptions in the Ultimate Heat Sink Inventory Calculation The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 27, 2011, the licensee failed to assure that the design basis information for expected heat loads to the ultimate heat sink was correctly translated into the ultimate heat sink 30-day inventory analysis. The analysis used a less conservative, frictionless form of the conservation of energy equation to determine heat load in the standby service water system during a 30-day design basis event. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07430 and CR-RBS-2011-07654.
The team determined that the failure to correctly translate expected heat loads into the ultimate heat sink inventory analysis was a performance deficiency. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to undesired consequences. Specifically, the neglect of friction heat load in the ultimate heat sink analysis system resulted in a condition where there was reasonable doubt on the operability of a system to meet its 30-day mission time without a makeup water source. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Specifically, the licensees revised analysis to determine operability removed overly conservative assumptions for operating the low pressure core spray pump for 30 days to account for the friction heat load added to the system. The finding has a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to thoroughly evaluate problems such that the resolutions address cause and extent of condition [P.1(c)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Residual Heat Removal Heat Exchanger Testing Frequency The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, from October 1998 to October 27, 2011, the licensee failed to establish a NRC Generic Letter 89-13 test program which incorporated a final test frequency for the residual heat removal heat exchangers and perform an adequate trending analysis upon which to base a final test frequency. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07713.
The team determined that failure to establish a NRC Generic Letter 89-13 test program which incorporated a final testing frequency of the residual heat removal heat exchangers was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inappropriate test frequency affected the licensees ability to ensure residual heat removal heat exchangers, when called upon, were available and capable to reliably perform as expected. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determine to have very low safety
 
significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding did not have a crosscutting aspect because the most significance contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Monitoring Standby Service Water System Leakage The team identified a Green, noncited violation of 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, which states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
Specifically, prior to October 27, 2011, the licensee failed to provide appropriate quantitative or qualitative acceptance criteria in station and abnormal operating procedures to determine if actions for leak detection were satisfactorily accomplished to protect the standby service water system and ultimate heat sink during design basis events. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07555.
The team determined that the failure to include appropriate acceptance criteria for leak detection in abnormal operating procedures for the standby service water system and ultimate heat sink was a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inadequate procedure guidance could lead to operators not recognizing conditions that would degrade the availability of the standby service water system. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 -
Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance: SL-IV Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for Change to Ultimate Heat Sink Inventory Requirements The team identified a Severity Level IV, noncited violation of 10 CFR 50.59, Changes, Tests and Experiments which states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if this activity would; result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the final safety analysis report (as updated). Specifically, from December 16, 2002, to October 27, 2011, the licensee changed the design basis of the ultimate heat sink inventory requirements to provide a 30-day cooling water supply without makeup capability to providing a less than 30-day cooling water supply with makeup capability without obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR 2011-07674.
The team determined that the failure to obtain a license amendment prior to implementing a proposed change, test or experiment to the ultimate heat sink requirements was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding has the ability to impact the regulatory process. The finding was more than minor because it involved a change to the updated final safety analysis report description where there was a reasonable likelihood that the change would require NRC approval. In accordance with the NRC Enforcement Policy, the team used insights from MC 0609, Significance Determination Process, to determine the final significance of the finding. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 -
 
Initial Screening and Characterization of Findings, the finding represented a loss of system safety function in that the ultimate heat sink could not meet its 30-day mission time to provide decay heat removal. Therefore, a Phase 2 evaluation was necessary. The significance of the finding could not be assessed quantitatively through a Phase 2 or Phase 3 analysis. Consequently, an assessment was performed in accordance with IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The finding was determined to have very low safety significance because the frequency of events that would require long term use of the ultimate heat sink is very low and the difference in the failure probability to replenish the ultimate heat sink in 10 days versus 30 days is very small. This was because an early depletion of the inventory would be easily detected and would become a priority. At the time that replenishment would be needed, plant conditions should be stable and local transportation arteries should be restored. Therefore, since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Abnormal Procedure for Reducing Loads on Standby Diesel Generators The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, and drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, prior to October 27, 2011, the licensee failed to include appropriate qualitative and quantitative acceptance criteria in abnormal operating procedures for control room operators to recognize the need to reduce loads on the standby diesel generators during design basis accidents. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07716.
The team determined that the failure to include appropriate quantitative or qualitative acceptance criteria in abnormal operating procedures for control room operators to recognize the need to reduce loads on the standby diesel generators during design basis accidents was a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, a control room operating crews failure to recognize the need to reduce loads to prevent the standby diesel generator failure during design basis accidents adversely affected the reliability of the standby diesel generators. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety for the correct training of licensed operator personnel [H.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency and Abnormal Procedures for Standby Diesel Generator Fail to Load Sequences The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, and drawings shall include appropriate qualitative and quantitative criteria for determining that important activities have been satisfactorily accomplished.
Specifically, prior to October 27, 2011, the licensee failed to include appropriate quantitative or qualitative acceptance criteria in procedures for control room operators to recognize and recover a standby diesel generator that starts but fails to load with the remaining standby diesel generator out of service during a loss-of-offsite-power event. This
 
finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07716, CR-RBS-2011-07717, and CR-RBS-2011-07718.
The team determined that the failure to include appropriate quantitative or qualitative acceptance criteria to determine that important activities are satisfactorily accomplished in emergency and abnormal operating procedures used during loss-of-offsite-power events was a performance deficiency. The finding was more than minor because it is associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, a control room operator crews failure to diagnose recoverable conditions adversely affected the availability of standby diesel generators during a loss-of-offsite-power event. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of problem identification and resolution, operating experience component, because the licensee did not implement and institutionalize operating experience through changes to station processes, procedures, equipment, and training programs [P.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Simulator Fidelity for Emergency Diesel Generator Loading The team identified a Green, noncited violation of 10 CFR 55.46(c)(1), Simulation Facilities, which states, in part, that a plant-referenced simulator used for the administration of the operating test must demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond. Specifically, prior to October 27, 2011, the River Bend Station simulator did not demonstrate the expected plant response for standby diesel generator loading during accident conditions to which the simulator was designed to respond. The electrical loading on the emergency diesel generator in the simulator was approximately 800 kW less than the expected full load for the diesel generator. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07682.
The team determined that the failure of the plant-referenced simulator to demonstrate expected plant response for standby diesel generator loading during accident conditions to which the simulator has been designed to respond was a performance deficiency. The finding was more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the incorrect simulator response adversely affected the control room operator crews capability to assess standby diesel generator loading conditions. In accordance with NRC Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets and the associated Appendix I, the finding was determined to be of very low safety significance (Green). Specifically, Manual Chapter 0609, Appendix I, Operator Requalification Human Performance Significance Determination Process, block 12, establishes a Green finding for failure to correctly replicate the plants response on the simulator that either has the potential to cause or actually causes negative training to operators. Negative training did occur for this finding because operators thought they had electrical load margin on the emergency diesel generators when the diesels were actually fully loaded with minimal margin without securing other equipment. This finding had a crosscutting aspect in the area of human performance, resources component, in that the licensee did not ensure that equipment (plant-referenced simulator) was adequate to assure nuclear safety for the correct training of licensed operator personnel [H.2(b)].
Inspection Report# : 2011008 (pdf)
Barrier Integrity
 
Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Maintenance Instructions used for Suppression Pool Cooling Isolation Valve Maintenance The inspectors identified a Green, self-revealing non-cited violation of Technical Specification 5.4.1.a, Procedures, for inadequate maintenance procedures to properly assemble containment isolation valves on the suppression pool cooling system. This resulted in a failure of the suppression pool cooling systems outboard containment isolation valve marriage coupling that ensures the valve stem is connected to the valve actuator. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2011-09171.
The failure to establish adequate work instructions to assemble the suppression pool cleanup system isolation valves is a performance deficiency. The inspectors determined that the finding was more than minor because it is associated with the Barrier Integrity Cornerstone attribute of Systems, Structures, and Components and Barrier Performance, and affected the cornerstone objective of providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations. Using the Phase 1 SDP worksheet for the barrier integrity cornerstone, the inspectors answered no to all four screening questions under the containment barrier column. Specifically, the affected penetration did not represent an actual open pathway in the physical integrity of reactor containment due to an operable and functionally redundant containment isolation valve in the suppression pool cooling piping penetration. The apparent cause of the finding was the failure of the planning department to recognize and develop design documentation to identify the set screw size and starting material necessary to determine the appropriate set screw torque for work affecting safety related equipment. The inspectors determined the finding had a cross cutting aspect in the human performance, area associated with the resources component because of the lack of complete accurate and up to date design documentation associated with the work package development. [H.2(c)].
Inspection Report# : 2012002 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Standby Gas Treatment Electric Heater Power Output Calculation The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, for an inadequate calculation methodology used in determining standby gas treatment system operability. The inspectors found that the calculation neither considered instrument uncertainty nor applied a proper voltage drop from the breaker to the standby gas treatment system filter train heater. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-07332.
The finding was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone to maintain radiological barrier functionality of standby gas treatment trains, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, operating the standby gas system filter train heaters without sufficient output power is detrimental to the charcoal filters ability to retain radioactive iodine. This could result in a greater amount of radiation release to the environment in the event of an accident. In accordance with Inspection manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the Phase 1 significance determination process screening determined the finding to be only of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the standby gas treatment system. The apparent cause of this finding was the decision to develop an engineering evaluation that did not include instrument uncertainly and did not validate the correct voltage drop between the filter train heater feeder breaker and the heater elements. The finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel failed to use conservative assumptions when developing the modified output power methodology for operation of the standby gas treatment system filter heaters with only 8 of 9 heater elements installed [H.1(b)].
Inspection Report# : 2011004 (pdf)
 
Emergency Preparedness Occupational Radiation Safety Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform a Radiation Survey Inspectors reviewed a self-revealing non-cited violation of 10 CFR 20.1501(a) for the failure to perform a radiation survey. A survey was not completed after two contaminated valves were transferred from the 98-foot elevation of the main steam tunnel to the radwaste area. During shift turnovers, workers responsible for transferring the valves did not understand that they needed to remove two buckets, and perform a survey after completing the valve transfer.
Consequently, a bucket with highly contaminated water and residual was left in the tunnel causing radiation levels as high as 300 millirem per hour. This resulted in an unposted high radiation area. The licensee entered the issue into the corrective actions program as Condition Report CR-RBS-2011-01552.
The failure to perform a radiation survey to evaluate the radiological conditions is a performance deficiency. The finding is more than minor because it negatively impacted the Occupational Radiation Safety cornerstones attribute of program and process, in that the lack of a post-work survey did not ensure exposure control for workers. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of 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 the ability to assess dose was not compromised. The finding has a Human Performance cross-cutting component associated with the aspect of work practices because expectations regarding procedural compliance for post-job radiation surveys were ineffective [H.4(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Control Access to a High Radiation Area Inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.7.1(c), resulting from the licensees failure to control access to a high radiation area. Specifically, a carpenter entered a high radiation area in the main steam tunnel near valve V112 without proper authorization before a health physics technician completed radiation surveys and received an unexpected alarming dosimeter reading of 110 millirem per hour. The carpenter had not been briefed that dose rates in the area measured 140 millirem per hour. He had been instructed not to perform any work before the health physics technician surveyed the area, but River Bend did not make it clear enough that he was to follow all health physics instructions. The licensee entered the issue into the corrective actions program as Condition Report CR-RBS-2011-01426 and the worker was counseled.
The failure to control access to a high radiation area was a performance deficiency. The finding was more than minor because it was associated with the occupational radiation safety attribute of exposure control and affected the cornerstone objective in that not controlling a high radiation area could increase personal exposure. In addition, this type of issue is addressed in Example 6.h of IMC 0612, Appendix E, Examples of Minor Issues. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The finding has a Human Performance cross-cutting component associated with the aspect of work practices because expectations regarding supervisory and management oversight of work activities, including
 
contractors to ensure that safety is supported were not met [H.4(c)].
Inspection Report# : 2012002 (pdf)
Public Radiation Safety Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Digital Radiation Monitoring System The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of the digital radiation monitoring system. Specifically, the maintenance rule expert panel performed an inadequate analysis after the digital radiation monitoring system exceeded the condition monitoring criteria by failing to follow the procedural requirements of EN-DC-206 to have cause evaluations for system failures so that maintenance preventability could be properly evaluated. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2011-00485.
The inspectors determined that the failure to adequately monitor the performance of the digital radiation monitoring system is a performance deficiency. The inspectors reviewed Inspection Manual Chapter (IMC) 0612 and determined that the finding is more than minor because the finding is associated with the plant facilities/equipment and instrumentation attribute (reliability of process radiation monitors) of the radiation safety cornerstone (public radiation safety) and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian use. The finding was assessed using the IMC 0609, Appendix D, Public Radiation SDP, and because there was no failure to implement the effluent program, the finding was determined to be of very low safety significance (Green). The inspectors reviewed the apparent cause of this finding and found that the oversight of the maintenance rule program was adversely affected by personnel changes and lack of effective turnover. Therefore, the finding has a cross-cutting aspect in the human performance area and resources component because the licensee failed to ensure that maintenance rule program personnel were trained and sufficiently qualified to perform their duties in an effective manner [H.2(b)].
Inspection Report# : 2012002 (pdf)
Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : September 12, 2012
 
3Q/2012 Inspection Findings - River Bend 1 River Bend 1 3Q/2012 Plant Inspection Findings Initiating Events Significance:        Jun 29, 2012 Identified By: NRC Item Type: FIN Finding Failure to Follow Procedure to Protect Sensitive Plant Areas The inspectors identified a finding for failure to follow Operating System Procedure OSP-0048, "Switchyard, Transformer Yard, and Sensitive Equipment Controls." Specifically, the licensee failed to appropriately consider the plant impact when planning and approving work in the main transformer yard and switchyard potentially introducing unacceptable risk to plant operations contrary to OSP-0048 administrative controls. This issue was entered into the licensees corrective action program as Condition Reports CR-RBS-2012-02479, CR-RBS-2012-02821, and CR-RBS-2012-04129.
The finding was more than minor in accordance with Appendix B, "Issue Screening," of Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," because the finding was associated with the protection against external events attribute of the Initiating Events cornerstone and affected 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. Specifically, the routine failure to integrate switchyard and transformer yard work into the River Bend work process increased the likelihood that unintended, uncoordinated maintenance and test activities could reduce the diversity of electrical power and cause inadvertent reductions in nuclear plant defense-in-depth. The inspectors performed a Phase 1 significance determination process review of this finding per Inspection Manual Chapter 0609, , "Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) since the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, nor did it contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, and the finding did not increase the likelihood of a fire or internal or external flooding. The inspectors determined the apparent cause of this finding was a lack of management oversight of station work activities. Therefore, this finding has a cross-cutting aspect in the area of human performance associated with the work practices component because station management failed to provide proper oversight of the process to protect sensitive areas of the plant [H.4(c)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: FIN Finding Failure to Implement Severe Weather Operations Procedure The inspectors identified a finding that involved failure to implement a procedure to protect the plant during adverse weather conditions. Specifically, appropriate equipment walkdowns and corrective actions were not performed to protect equipment important to safety from severe weather risks in a timely manner. The concerns were documented in Condition Report CR-RBS-2012-02387.
The finding was determined to be of very low safety significance (Green) since the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, nor did it contribute to both the Page 1 of 15
 
3Q/2012 Inspection Findings - River Bend 1 likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, and the finding did not increase the likelihood of a fire or internal or external flooding. The inspectors determined the apparent cause of this finding was operations expectation that excellent housekeeping nominally exists in the switchyard and transformer yard. Therefore, there was no need to dispatch personnel to verify housekeeping because that action would risk personnel safety. The status of an unsecured ladder in the transformer yard is evidence that up to date information is essential to confirm whether housekeeping is satisfactory. Therefore, the finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the station did not demonstrate that nuclear safety was an overriding priority because it failed to implement the roles and authorities in their severe weather operations procedure [H.1(a)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Properly Assemble Turbine Control Valve Push Rod-Spring Housing Coupling The inspectors reviewed a self-revealing finding associated with main turbine control valve number 3 unexpectedly closing. In response, operators reduced reactor power to 90 percent. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-02773.
The finding was more than minor because it was associated with the Initiating Events cornerstone attribute of design control and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability by resulting in a plant downpower and subsequent planned outage for repair activities. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affecting mitigating equipment, and did not increase the likelihood of a fire or flood. The inspectors did not identify a cross cutting aspect because the performance deficiency is not indicative of the licensees current performance.
Inspection Report# : 2012003 (pdf)
Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Relief Valve Configuration Control Results in a Reactor Downpower The inspectors identified a self-revealing finding for failing to maintain configuration control of the gland seal header relief valves bonnet vent port. The configuration control failure lead to a subsequent decrease in condenser vacuum requiring an unplanned power reduction to maintain adequate condenser vacuum margin. This finding has been entered into the licensees corrective action program as Condition Report CR-RBS-2012-00736.
The failure to maintain configuration control of the glad seal header relief valve was a performance deficiency. The finding was determined to be more than minor because it was associated with the configuration control attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the failure to maintain configuration control resulted in an unplanned down power. Using Inspection Manual Chapter IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems will not be available. The inspectors determined that the apparent cause of this finding was that when the licensee prepared work orders that directed installation of the gland seal header relief valves, they did not comply with procedural requirements to provide plant configuration controls. Therefore, this Page 2 of 15
 
3Q/2012 Inspection Findings - River Bend 1 finding has a cross-cutting aspect in the human performance area associated with the work practice component because the licensee did not define and effectively communicate expectations regarding procedural compliance [H.4 (b)].
Inspection Report# : 2012002 (pdf)
Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Properly Fabricate and Install the mid-Standard Turbine Shaft Brush The inspectors reviewed a self-revealing finding regarding the improper fabrication of a turbine shaft grounding brush that resulted in turbine trip and subsequent reactor scram. The licensee identified the improper fabrication of a turbine shaft grounding brush as the cause of a spurious main turbine over-speed trip signal from an electrical discharge from the turbine shaft. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-9053.
Failure to fabricate the turbine shaft grounding brush in accordance with vendor instructions is a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. Specifically, the improperly fabricated grounding brush resulted in a turbine trip and subsequent reactor scram. The inspectors reviewed the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affecting mitigating equipment, and did not increase the likelihood of a fire or flood. The apparent cause of the performance deficiency was the failure in 2004 to appropriately perform a post maintenance test for the turbine shaft grounding brush modification. Therefore the inspectors did not identify a cross-cutting aspect because the performance deficiency is not reflective of the licensees current performance.
Inspection Report# : 2012002 (pdf)
Mitigating Systems Significance:        Jun 29, 2012 Identified By: NRC Item Type: NCV NonCited Violation High Pressure Core Spray Diesel Generator Bearing Lubrication Deficiencies The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failing to correct a condition adverse to quality for lubricating the high pressure core spray diesel generator bearings.
The station documented the finding in Condition Report CR-RBS-2012-02666.
This performance deficiency was more than minor and was a finding because, if left uncorrected, inadequate lubrication work instruction could cause bearing failure due to inadequate lubrication or generator winding failure due to grease intrusion into the electrical windings in the generator. The significance of this finding was evaluated using a Phase 1 significance determination process screening and was determined to be of 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 seismic, flooding, or severe weather initiating events. The apparent Page 3 of 15
 
3Q/2012 Inspection Findings - River Bend 1 reason the initial condition report was closed without correcting the work instruction to lubricate the high pressure core spray diesel generator bearings was that personnel who prepared and approved the operability evaluation were focused on proving operability not correcting a condition adverse to quality. Their focus was specific to the components ability to perform its function and not on completely identifying the issue in the corrective action program. Therefore, the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the station did not identify this issue completely, accurately, and in a timely manner commensurate with its safety significance [P.1(a)].
Inspection Report# : 2012003 (pdf)
Significance:      Jun 29, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Specify Manual Actions for Safety Relief Valve Operations During a Station Blackout Event The inspectors identified a non-cited violation of 10 CFR 50.63, Loss of All Alternating Current, paragraph (a) (2),
which states, in part, The reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. The capability for coping with a station blackout of specified duration shall be determined by an appropriate coping analysis. Licensees are expected to have the baseline assumptions, analyses, and related information used in their coping evaluations available for NRC review. Specifically, from November 1985 to May 17, 2012, the licensee failed to specify actions while ac power is unavailable to ensure that safety relief valves provided sufficient capacity and capability to ensure appropriate containment integrity is maintained during a station blackout event. This violation has been entered into the corrective action program as Condition Report CR-RBS-2012-03376.
The inspectors determined that failure to specify actions for safety relief valve operation in procedures in accordance with NUMARC-8700 was a performance deficiency. The finding was more than minor because it adversely affected the procedure quality attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to respond to undesirable consequences. Specifically, the station blackout coping procedures did not specify actions that would ensure the heat capacity temperature limit for the suppression pool would not be exceeded during the station blackout coping period. Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the inspectors determined that the Mitigating Systems Cornerstone was affected because the finding could cause degradation of core decay heat removal. Using Table 4a from the Phase 1 worksheet, the inspectors determined that the finding represents a loss of safety function; therefore, a Phase 2 analysis was necessary. However, the inspectors determined that a Phase 2 analysis was not sufficient to assess significance because of the complexity of the finding. Therefore, a Phase 3 analysis was necessary. The result of the Phase 3 analysis determined that the change in core-damage-frequency (?CDF) for the performance deficiency was 2.4E-7 or very low safety significance (Green). The senior reactor analyst determined that the change in large-early-release-frequency (?LERF) was 4.8E-8 or very low safety significance (Green). No cross-cutting aspect was identified because the most significant contributor was not indicative of current licensee performance (Section 4OA5).
Inspection Report# : 2012003 (pdf)
Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Appropriately Assess and Manage Risk for Internal Flooding Events 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, due to the failure of work control and operations personnel to adequately assess the increase in risk associated with internal flooding events. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2012-00641.
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3Q/2012 Inspection Findings - River Bend 1 The failure of work control and operations personnel to adequately assess the risk associated with internal flooding is a performance deficiency. The performance deficiency resulted in the overall elevated plant risk placing the plant into the higher licensee-established risk category (Green to Yellow). The performance deficiency 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. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowcharts 1 and 2, the finding was determined to have very low safety significance (Green) because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability deficit was less than 1E-7. The inspectors determined that the apparent cause of the finding was that station personnel routinely failed to review the qualitative risk checklist required by the stations risk management procedure. Therefore, this finding has a cross-cutting aspect in the human performance area associated with the work practice component because the licensee did not define and effectively communicate expectations regarding procedural compliance. [H.4(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Assumptions used in Standby Equipment Room Temperature Analysis The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because, prior to February 7, 2012, the licensee did not verify that assumptions used in confirming that the safety-related battery inverter rooms would remain below their design basis temperature limits during a design basis event agreed with the as-built condition of the plant. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2012-01046.
The inspectors determined that the failure to verify that design documents match the actual configuration of the plant is a performance deficiency. The finding was more than minor because it adversely affects the Mitigating Systems Cornerstone objective of equipment performance to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee had not verified assumptions that ensure the standby switchgear room air conditioning system would reliably maintain the standby equipment rooms below the design temperature limits. Using Inspection Manual Chapter 0609, Attachment 4, "Initial Screening and Characterization of Findings," the finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function, nor actual loss of safety function of a single train, and it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that this issue has a cross-cutting aspect in the area of human performance decision-making regarding nonconservative assumptions. When the licensee conducted the flow balance test, they assumed that measuring air inflow alone was sufficient, but did not check that the doors gaps were allowing a sufficient amount of warm air to exit standby equipment rooms and be circulated back to the general areas [H.1(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Appropriately Set Reactor Core Isolation Cooling Flow Controller High Output Limit The inspectors identified a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Procedures, when the reactor core isolation cooling turbine tripped on mechanical over speed. Troubleshooting determined the cause was an improperly tuned flow controller. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2012-01188 and CR-RBS-2012-01262.
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3Q/2012 Inspection Findings - River Bend 1 The failure to provide specific flow controller tuning instructions for the reactor core isolation cooling turbine flow controller was a performance deficiency. The finding was more than minor in accordance with Appendix B, "Issue Screening," of Inspection Manual Chapter IMC 0612, "Power Reactor Inspection Reports," because the finding was associated with the equipment performance 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, improper tuning of the reactor core isolation cooling controller impacted operability and availability of the reactor core isolation cooling system. The inspectors performed a Phase 1 significance determination process review of this finding per Inspection Manual Chapter IMC 0609, Attachment 4, "Initial Screening and Characterization of Findings." In accordance with Table 4a, "Characterization Worksheet for IE, MS, and BI Cornerstones," the finding represented a loss of system safety function. Therefore, a Region IV senior reactor analyst used Inspection Manual Chapter IMC 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," to review the finding using the Standardized Plant Analysis Risk (SPAR) model for River Bend Station. The Phase 3 analysis determined the Delta-CDF was 4.68E-7/yr. For a 7-month exposure, the incremental conditional core damage probability is 2.73E-7. The majority of the risk came from sequences involving a loss of feedwater (48 percent) and a loss of offsite power (33 percent). Consequently, the analyst determined that the risk associated with the performance deficiency was very low (green). The inspectors determined the apparent cause of this finding was the failure to perform a post maintenance test to identify that the high output limit was not properly set by the maintenance work instruction. Therefore, this finding has cross-cutting aspect in the area of human performance associated with the resources component due to less than adequate work package testing instruction. [H.2(c)].
Inspection Report# : 2012002 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Standby Service Water Pump Motor Lubrication Deficiencies The inspectors identified a Green non-cited violation of Technical Specification 5.4.1.a, because the station did not establish appropriate maintenance procedures to lubricate standby service water pump lower motor bearings.
Specifically, the inspectors found a legacy of improper maintenance practices involving lubrication of the standby service water pump motor lower bearings going back to 1986. This included mixing of incompatible greases without change evaluations, lubrication techniques that did not comply with pump motor vendor manual or EPRI guidance, improper volume of greases added to the bearings, and improper preventive maintenance frequency for performing re-greasing of the bearings. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-08367.
This performance deficiency is more-than-minor and is therefore a finding because if left uncorrected, this performance deficiency has the potential to lead to a more significant safety concern. Specifically, if the subject work orders are not corrected, future work activities that grease the subject bearings in accordance with those work orders may not grease the bearings adequately, which may result in common-cause failures of the station service water pumps. Because this finding was identified while the unit was operating, the inspectors used MC 0609 Appendix A to assess its risk significance. In accordance with that Appendix, the finding screened as green (of very low safety significance) because it was not a design or qualification deficiency; it did not represent a loss of system safety function; and it did not screen as potentially risk-significant due to seismic, flooding, or severe weather initiating events. The inspectors determined that the apparent cause of this finding was failure to include the appropriate scope of information in the work instructions due to overconfidence and lack of adequate review by engineering staff.
Specifically, the system engineer who developed the revised instructions failed to develop appropriate steps with adequate detail to appropriately perform the task and the field engineer failed to stop work and discuss the issue with the system engineer that developed the work instructions. Therefore, the finding has a crosscutting aspect in the area of human performance associated with work practices, because engineering personnel failed to use the applicable Page 6 of 15
 
3Q/2012 Inspection Findings - River Bend 1 human error prevention techniques [H.4(a)].
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Control Building Chiller System The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of the control building chilled water system. Specifically, the inspectors determined that the station had failed to track system unavailability following the systems classification of a high risk system and did not monitor the system at the train level, ultimately masking the performance of individual trains. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-07332.
The finding was more than minor since violations of 10 CFR 50.65(a)(2) necessarily involve degraded system performance which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the finding did not lead to an actual loss of safety function of the system or cause a component to be inoperable, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the cause of the finding was the lack of management oversight.
Following the issuance of River Bend Station Probabilistic Risk Assessment interim Revision 4a, several personnel functioned as the maintenance rule coordinator and control building chilled water system engineer. During this period, station management did not ensure sufficient knowledge transfer for effective maintenance rule implementation.
Therefore, this finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to ensure supervisory and management oversight of work activities such that nuclear safety is supported [H.4(c)].
Inspection Report# : 2011005 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Testing of Division I and Division III Standby Diesel Generators The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, prior to October 27, 2011, the licensee failed to ensure surveillance testing procedures of Division I and III standby diesel generators incorporated the correct acceptance limits for maximum expected load at max frequency and voltage specified in design basis documents. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07132, CR-RBS-2011-07294, and CR-RBS-2011-07518.
The team determined that the failure to ensure that the test procedures required to demonstrate that Division I and Division III standby diesel generators will perform satisfactorily in service incorporated the requirements and acceptance limits contained in applicable design documents was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee could not ensure that the standby diesel generators would reliably provide power for the maximum expected post-accident loads including maximum frequency and voltage. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1
- Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or Page 7 of 15
 
3Q/2012 Inspection Findings - River Bend 1 functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of condition
[P.1(c)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Conservative Design Assumptions in the Ultimate Heat Sink Inventory Calculation The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 27, 2011, the licensee failed to assure that the design basis information for expected heat loads to the ultimate heat sink was correctly translated into the ultimate heat sink 30-day inventory analysis. The analysis used a less conservative, frictionless form of the conservation of energy equation to determine heat load in the standby service water system during a 30-day design basis event. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07430 and CR-RBS-2011-07654.
The team determined that the failure to correctly translate expected heat loads into the ultimate heat sink inventory analysis was a performance deficiency. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to undesired consequences. Specifically, the neglect of friction heat load in the ultimate heat sink analysis system resulted in a condition where there was reasonable doubt on the operability of a system to meet its 30-day mission time without a makeup water source. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Specifically, the licensees revised analysis to determine operability removed overly conservative assumptions for operating the low pressure core spray pump for 30 days to account for the friction heat load added to the system. The finding has a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to thoroughly evaluate problems such that the resolutions address cause and extent of condition [P.1(c)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Residual Heat Removal Heat Exchanger Testing Frequency The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, from October 1998 to October 27, 2011, the licensee failed to establish a NRC Generic Letter 89-13 test program which incorporated a final test frequency for the residual heat removal heat exchangers and perform an Page 8 of 15
 
3Q/2012 Inspection Findings - River Bend 1 adequate trending analysis upon which to base a final test frequency. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07713.
The team determined that failure to establish a NRC Generic Letter 89-13 test program which incorporated a final testing frequency of the residual heat removal heat exchangers was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inappropriate test frequency affected the licensees ability to ensure residual heat removal heat exchangers, when called upon, were available and capable to reliably perform as expected. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determine to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding did not have a crosscutting aspect because the most significance contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Monitoring Standby Service Water System Leakage The team identified a Green, noncited violation of 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, which states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
Specifically, prior to October 27, 2011, the licensee failed to provide appropriate quantitative or qualitative acceptance criteria in station and abnormal operating procedures to determine if actions for leak detection were satisfactorily accomplished to protect the standby service water system and ultimate heat sink during design basis events. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07555.
The team determined that the failure to include appropriate acceptance criteria for leak detection in abnormal operating procedures for the standby service water system and ultimate heat sink was a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the inadequate procedure guidance could lead to operators not recognizing conditions that would degrade the availability of the standby service water system. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 -
Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance: SL-IV Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for Change to Ultimate Heat Sink Inventory Requirements The team identified a Severity Level IV, noncited violation of 10 CFR 50.59, Changes, Tests and Experiments Page 9 of 15
 
3Q/2012 Inspection Findings - River Bend 1 which states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if this activity would; result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the final safety analysis report (as updated). Specifically, from December 16, 2002, to October 27, 2011, the licensee changed the design basis of the ultimate heat sink inventory requirements to provide a 30-day cooling water supply without makeup capability to providing a less than 30-day cooling water supply with makeup capability without obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR 2011-07674.
The team determined that the failure to obtain a license amendment prior to implementing a proposed change, test or experiment to the ultimate heat sink requirements was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding has the ability to impact the regulatory process. The finding was more than minor because it involved a change to the updated final safety analysis report description where there was a reasonable likelihood that the change would require NRC approval. In accordance with the NRC Enforcement Policy, the team used insights from MC 0609, Significance Determination Process, to determine the final significance of the finding. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 -
Initial Screening and Characterization of Findings, the finding represented a loss of system safety function in that the ultimate heat sink could not meet its 30-day mission time to provide decay heat removal. Therefore, a Phase 2 evaluation was necessary. The significance of the finding could not be assessed quantitatively through a Phase 2 or Phase 3 analysis. Consequently, an assessment was performed in accordance with IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The finding was determined to have very low safety significance because the frequency of events that would require long term use of the ultimate heat sink is very low and the difference in the failure probability to replenish the ultimate heat sink in 10 days versus 30 days is very small. This was because an early depletion of the inventory would be easily detected and would become a priority. At the time that replenishment would be needed, plant conditions should be stable and local transportation arteries should be restored. Therefore, since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Abnormal Procedure for Reducing Loads on Standby Diesel Generators The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, and drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, prior to October 27, 2011, the licensee failed to include appropriate qualitative and quantitative acceptance criteria in abnormal operating procedures for control room operators to recognize the need to reduce loads on the standby diesel generators during design basis accidents. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07716.
The team determined that the failure to include appropriate quantitative or qualitative acceptance criteria in abnormal operating procedures for control room operators to recognize the need to reduce loads on the standby diesel generators during design basis accidents was a performance deficiency. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, a control room operating crews failure to recognize the need to reduce loads to prevent the standby diesel generator failure during design basis accidents adversely affected the reliability of the standby diesel generators. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Page 10 of 15
 
3Q/2012 Inspection Findings - River Bend 1 Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety for the correct training of licensed operator personnel [H.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency and Abnormal Procedures for Standby Diesel Generator Fail to Load Sequences The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, and drawings shall include appropriate qualitative and quantitative criteria for determining that important activities have been satisfactorily accomplished.
Specifically, prior to October 27, 2011, the licensee failed to include appropriate quantitative or qualitative acceptance criteria in procedures for control room operators to recognize and recover a standby diesel generator that starts but fails to load with the remaining standby diesel generator out of service during a loss-of-offsite-power event. This finding was entered into the licensees corrective action program as Condition Reports CR-RBS-2011-07716, CR-RBS-2011-07717, and CR-RBS-2011-07718.
The team determined that the failure to include appropriate quantitative or qualitative acceptance criteria to determine that important activities are satisfactorily accomplished in emergency and abnormal operating procedures used during loss-of-offsite-power events was a performance deficiency. The finding was more than minor because it is associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, a control room operator crews failure to diagnose recoverable conditions adversely affected the availability of standby diesel generators during a loss-of-offsite-power event. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability or functionality, loss of a system safety function, loss of a single train for greater than technical specification allowed outage time, loss of one or more non-technical specification risk significant equipment for greater than 24 hours, and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of problem identification and resolution, operating experience component, because the licensee did not implement and institutionalize operating experience through changes to station processes, procedures, equipment, and training programs [P.2(b)].
Inspection Report# : 2011008 (pdf)
Significance:        Oct 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Simulator Fidelity for Emergency Diesel Generator Loading The team identified a Green, noncited violation of 10 CFR 55.46(c)(1), Simulation Facilities, which states, in part, that a plant-referenced simulator used for the administration of the operating test must demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond. Specifically, prior to October 27, 2011, the River Bend Station simulator did not demonstrate the Page 11 of 15
 
3Q/2012 Inspection Findings - River Bend 1 expected plant response for standby diesel generator loading during accident conditions to which the simulator was designed to respond. The electrical loading on the emergency diesel generator in the simulator was approximately 800 kW less than the expected full load for the diesel generator. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2011-07682.
The team determined that the failure of the plant-referenced simulator to demonstrate expected plant response for standby diesel generator loading during accident conditions to which the simulator has been designed to respond was a performance deficiency. The finding was more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the incorrect simulator response adversely affected the control room operator crews capability to assess standby diesel generator loading conditions. In accordance with NRC Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets and the associated Appendix I, the finding was determined to be of very low safety significance (Green). Specifically, Manual Chapter 0609, Appendix I, Operator Requalification Human Performance Significance Determination Process, block 12, establishes a Green finding for failure to correctly replicate the plants response on the simulator that either has the potential to cause or actually causes negative training to operators. Negative training did occur for this finding because operators thought they had electrical load margin on the emergency diesel generators when the diesels were actually fully loaded with minimal margin without securing other equipment. This finding had a crosscutting aspect in the area of human performance, resources component, in that the licensee did not ensure that equipment (plant-referenced simulator) was adequate to assure nuclear safety for the correct training of licensed operator personnel [H.2(b)].
Inspection Report# : 2011008 (pdf)
Barrier Integrity Significance:        Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Maintenance Instructions used for Suppression Pool Cooling Isolation Valve Maintenance The inspectors identified a Green, self-revealing non-cited violation of Technical Specification 5.4.1.a, Procedures, for inadequate maintenance procedures to properly assemble containment isolation valves on the suppression pool cooling system. This resulted in a failure of the suppression pool cooling systems outboard containment isolation valve marriage coupling that ensures the valve stem is connected to the valve actuator. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2011-09171.
The failure to establish adequate work instructions to assemble the suppression pool cleanup system isolation valves is a performance deficiency. The inspectors determined that the finding was more than minor because it is associated with the Barrier Integrity Cornerstone attribute of Systems, Structures, and Components and Barrier Performance, and affected the cornerstone objective of providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations. Using the Phase 1 SDP worksheet for the barrier integrity cornerstone, the inspectors answered no to all four screening questions under the containment barrier column. Specifically, the affected penetration did not represent an actual open pathway in the physical integrity of reactor containment due to an operable and functionally redundant containment isolation valve in the suppression pool cooling piping penetration. The apparent cause of the finding was the failure of the planning department to recognize and develop design documentation to identify the set screw size and starting material necessary to determine the appropriate set screw torque for work affecting safety related Page 12 of 15
 
3Q/2012 Inspection Findings - River Bend 1 equipment. The inspectors determined the finding had a cross cutting aspect in the human performance, area associated with the resources component because of the lack of complete accurate and up to date design documentation associated with the work package development. [H.2(c)].
Inspection Report# : 2012002 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform a Radiation Survey Inspectors reviewed a self-revealing non-cited violation of 10 CFR 20.1501(a) for the failure to perform a radiation survey. A survey was not completed after two contaminated valves were transferred from the 98-foot elevation of the main steam tunnel to the radwaste area. During shift turnovers, workers responsible for transferring the valves did not understand that they needed to remove two buckets, and perform a survey after completing the valve transfer.
Consequently, a bucket with highly contaminated water and residual was left in the tunnel causing radiation levels as high as 300 millirem per hour. This resulted in an unposted high radiation area. The licensee entered the issue into the corrective actions program as Condition Report CR-RBS-2011-01552.
The failure to perform a radiation survey to evaluate the radiological conditions is a performance deficiency. The finding is more than minor because it negatively impacted the Occupational Radiation Safety cornerstones attribute of program and process, in that the lack of a post-work survey did not ensure exposure control for workers. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of 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 the ability to assess dose was not compromised. The finding has a Human Performance cross-cutting component associated with the aspect of work practices because expectations regarding procedural compliance for post-job radiation surveys were ineffective [H.4(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Control Access to a High Radiation Area Inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.7.1(c), resulting from the licensees failure to control access to a high radiation area. Specifically, a carpenter entered a high radiation area in the main steam tunnel near valve V112 without proper authorization before a health physics technician completed radiation surveys and received an unexpected alarming dosimeter reading of 110 millirem per hour. The carpenter had not been briefed that dose rates in the area measured 140 millirem per hour. He had been instructed not to perform any work before the health physics technician surveyed the area, but River Bend did not make it clear enough that he was to follow all health physics instructions. The licensee entered the issue into the corrective actions program as Condition Report CR-RBS-2011-01426 and the worker was counseled.
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3Q/2012 Inspection Findings - River Bend 1 The failure to control access to a high radiation area was a performance deficiency. The finding was more than minor because it was associated with the occupational radiation safety attribute of exposure control and affected the cornerstone objective in that not controlling a high radiation area could increase personal exposure. In addition, this type of issue is addressed in Example 6.h of IMC 0612, Appendix E, Examples of Minor Issues. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The finding has a Human Performance cross-cutting component associated with the aspect of work practices because expectations regarding supervisory and management oversight of work activities, including contractors to ensure that safety is supported were not met [H.4(c)].
Inspection Report# : 2012002 (pdf)
Public Radiation Safety Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Digital Radiation Monitoring System The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of the digital radiation monitoring system. Specifically, the maintenance rule expert panel performed an inadequate analysis after the digital radiation monitoring system exceeded the condition monitoring criteria by failing to follow the procedural requirements of EN-DC-206 to have cause evaluations for system failures so that maintenance preventability could be properly evaluated. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2011-00485.
The inspectors determined that the failure to adequately monitor the performance of the digital radiation monitoring system is a performance deficiency. The inspectors reviewed Inspection Manual Chapter (IMC) 0612 and determined that the finding is more than minor because the finding is associated with the plant facilities/equipment and instrumentation attribute (reliability of process radiation monitors) of the radiation safety cornerstone (public radiation safety) and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian use. The finding was assessed using the IMC 0609, Appendix D, Public Radiation SDP, and because there was no failure to implement the effluent program, the finding was determined to be of very low safety significance (Green). The inspectors reviewed the apparent cause of this finding and found that the oversight of the maintenance rule program was adversely affected by personnel changes and lack of effective turnover. Therefore, the finding has a cross-cutting aspect in the human performance area and resources component because the licensee failed to ensure that maintenance rule program personnel were trained and sufficiently qualified to perform their duties in an effective manner [H.2(b)].
Inspection Report# : 2012002 (pdf)
Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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3Q/2012 Inspection Findings - River Bend 1 Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Last modified : November 30, 2012 Page 15 of 15
 
4Q/2012 Inspection Findings - River Bend 1 River Bend 1 4Q/2012 Plant Inspection Findings Initiating Events Significance:        Oct 11, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Adequate Controlling Procedure for Stroking Safety Relief Valves at Low Power The inspectors identified a Green non-cited violation of Technical Specification 5.4.1.a for the failure to develop adequate controls for low-power stroking of safety relief valves. In response to this finding, the licensee trained senior reactor operators on the lessons learned from the finding. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-03816.
The performance deficiency was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Section B, this finding screened to very low safety significance because it was a transient initiator that did not result in a reactor trip and loss of mitigation equipment. Because the most significant causal factor of the performance deficiency was that the licensee had made an inappropriate assumption that the abnormal operating procedure was a satisfactory controlling document, this finding has a human performance cross cutting aspect associated with the decision making component, in that the licensee failed to use conservative assumptions in decision-making [H.1(b)] (Section 4OA5.3).
Inspection Report# : 2012010 (pdf)
Significance:        Oct 11, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Corrective Actions for Lockout Relay Failures The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality. Specifically, after a lockout relay mechanically bound in 2011, causing a fire, the licensee failed to identify and correct other susceptible relays. In response, the licensee tested other susceptible relays and replaced those that failed the test. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-05894.
This performance deficiency was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Section B, this finding screened to a detailed risk evaluation because it had caused a reactor trip and the loss of mitigation equipment such as loss of main feedwater and normal service water. The detailed risk evaluation included a quantitative bounding analysis and a qualitative evaluation in accordance with NRC Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, to determine that this finding was of very low safety significance (Green). Because the most significant causal factor of the performance deficiency was that the licensee had failed to recognize the potential risk to the plant when performing the evaluations for the failed lockout relays, this finding has a human performance cross-cutting aspect associated with the work control component in that licensee did not plan and coordinate work activities by incorporating risk insights, consistent with nuclear safety [H.3(a)] (Section 4OA5.4).
Inspection Report# : 2012010 (pdf)
Page 1 of 15
 
4Q/2012 Inspection Findings - River Bend 1 Significance:        Oct 11, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Test Lockout Relays in Accordance with Vendor Testing Practices The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, for the licensees failure to establish adequate preventative maintenance instructions for lockout relays in accordance with vendor recommendations for electrical testing. In response, the licensee incorporated vendor recommendations into the instructions for testing lockout relays. The licensee entered the finding into the corrective action program as Condition Report CR RBS-2011-02209.
The performance deficiency was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, in that it resulted in a fire. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Section B, this finding screened to very low safety significance (Green) because it was a transient initiator that did not result in a reactor trip or loss of mitigation equipment. The finding did not have a cross-cutting aspect because the performance deficiency was not representative of current plant performance (Section 4OA5.5).
Inspection Report# : 2012010 (pdf)
Significance:        Oct 11, 2012 Identified By: NRC Item Type: FIN Finding Failure to Establish An Adequate Cable Reliability Program The inspectors reviewed a self-revealing finding for the licensees failure to establish an effective cable reliability program, in that the licensee failed to distinguish between wetted and dry splices. In response, the licensee tested the high-risk-ranked cables, and replaced those that failed the test. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-03440.
The performance deficiency was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, in that it resulted in a reactor scram. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Section B, this finding screened to very low safety significance (Green) because it was a transient initiator that did not result in both a reactor trip and loss of mitigation equipment. Because the most significant causal factor of the performance deficiency was that the licensee failed to implement and institutionalize operating experience related to wetted splices, this finding has a problem identification and resolution cross cutting aspect associated with operating experience in that the licensee did not implement and institutionalize operating experience through changes to station processes and procedures to support plant safety [P.2 (b)] (Section 4OA5.6).
Inspection Report# : 2012010 (pdf)
Significance:        Sep 28, 2012 Identified By: NRC Item Type: FIN Finding Inadequate Verification of Leading Edge Flow Meter Functionality The inspectors identified a finding for the licensees failure to calibrate the feed water Leading Edge Flow Meter (LEFM) CheckPlus System following maintenance activities. This resulted in an error in reactor feed water flow rate data used to calculate reactor core thermal power. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-06274.
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4Q/2012 Inspection Findings - River Bend 1 This performance deficiency is more-than-minor and is therefore a finding because it was associated with the procedure quality attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The performance deficiency challenged the initiating events cornerstone objective by allowing the licensee to operate the plant outside of the prescribed analyzed uncertainty value, used in determining maximum core thermal power. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that this finding has very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the apparent cause of this finding was that when the licensee had changed the flow meter maintenance work scope that required transducer replacement, they had not included the vendor verification requirement in the revised work order. Therefore, this finding has a cross-cutting aspect in the Human Performance area of Work Control because the licensee had failed to appropriately coordinate the impact of changes to the work scope or activity on the plant. [H.3(b)].
Inspection Report# : 2012004 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: FIN Finding Failure to Follow Procedure to Protect Sensitive Plant Areas The inspectors identified a finding for failure to follow Operating System Procedure OSP-0048, "Switchyard, Transformer Yard, and Sensitive Equipment Controls." Specifically, the licensee failed to appropriately consider the plant impact when planning and approving work in the main transformer yard and switchyard potentially introducing unacceptable risk to plant operations contrary to OSP-0048 administrative controls. This issue was entered into the licensees corrective action program as Condition Reports CR-RBS-2012-02479, CR-RBS-2012-02821, and CR-RBS-2012-04129.
The finding was more than minor in accordance with Appendix B, "Issue Screening," of Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," because the finding was associated with the protection against external events attribute of the Initiating Events cornerstone and affected 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. Specifically, the routine failure to integrate switchyard and transformer yard work into the River Bend work process increased the likelihood that unintended, uncoordinated maintenance and test activities could reduce the diversity of electrical power and cause inadvertent reductions in nuclear plant defense-in-depth. The inspectors performed a Phase 1 significance determination process review of this finding per Inspection Manual Chapter 0609, , "Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) since the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, nor did it contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, and the finding did not increase the likelihood of a fire or internal or external flooding. The inspectors determined the apparent cause of this finding was a lack of management oversight of station work activities. Therefore, this finding has a cross-cutting aspect in the area of human performance associated with the work practices component because station management failed to provide proper oversight of the process to protect sensitive areas of the plant [H.4(c)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: FIN Finding Failure to Implement Severe Weather Operations Procedure The inspectors identified a finding that involved failure to implement a procedure to protect the plant during adverse weather conditions. Specifically, appropriate equipment walkdowns and corrective actions were not performed to protect equipment important to safety from severe weather risks in a timely manner. The concerns were documented in Condition Report CR-RBS-2012-02387.
The finding was determined to be of very low safety significance (Green) since the finding did not contribute to the Page 3 of 15
 
4Q/2012 Inspection Findings - River Bend 1 likelihood of a primary or secondary system loss of coolant accident initiator, nor did it contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, and the finding did not increase the likelihood of a fire or internal or external flooding. The inspectors determined the apparent cause of this finding was operations expectation that excellent housekeeping nominally exists in the switchyard and transformer yard. Therefore, there was no need to dispatch personnel to verify housekeeping because that action would risk personnel safety. The status of an unsecured ladder in the transformer yard is evidence that up to date information is essential to confirm whether housekeeping is satisfactory. Therefore, the finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the station did not demonstrate that nuclear safety was an overriding priority because it failed to implement the roles and authorities in their severe weather operations procedure [H.1(a)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Properly Assemble Turbine Control Valve Push Rod-Spring Housing Coupling The inspectors reviewed a self-revealing finding associated with main turbine control valve number 3 unexpectedly closing. In response, operators reduced reactor power to 90 percent. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-02773.
The finding was more than minor because it was associated with the Initiating Events cornerstone attribute of design control and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability by resulting in a plant downpower and subsequent planned outage for repair activities. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affecting mitigating equipment, and did not increase the likelihood of a fire or flood. The inspectors did not identify a cross cutting aspect because the performance deficiency is not indicative of the licensees current performance.
Inspection Report# : 2012003 (pdf)
Significance:        Mar 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Relief Valve Configuration Control Results in a Reactor Downpower The inspectors identified a self-revealing finding for failing to maintain configuration control of the gland seal header relief valves bonnet vent port. The configuration control failure lead to a subsequent decrease in condenser vacuum requiring an unplanned power reduction to maintain adequate condenser vacuum margin. This finding has been entered into the licensees corrective action program as Condition Report CR-RBS-2012-00736.
The failure to maintain configuration control of the glad seal header relief valve was a performance deficiency. The finding was determined to be more than minor because it was associated with the configuration control attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the failure to maintain configuration control resulted in an unplanned down power. Using Inspection Manual Chapter IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems will not be available. The inspectors determined that the apparent cause of this finding was that when the licensee prepared work orders that directed installation of the gland seal header relief valves, they did not comply with procedural requirements to provide plant configuration controls. Therefore, this finding has a cross-cutting aspect in the human performance area associated with the work practice component because the licensee did not define and effectively communicate expectations regarding procedural compliance [H.4 (b)].
Inspection Report# : 2012002 (pdf)
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4Q/2012 Inspection Findings - River Bend 1 Significance:        Mar 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Properly Fabricate and Install the mid-Standard Turbine Shaft Brush The inspectors reviewed a self-revealing finding regarding the improper fabrication of a turbine shaft grounding brush that resulted in turbine trip and subsequent reactor scram. The licensee identified the improper fabrication of a turbine shaft grounding brush as the cause of a spurious main turbine over-speed trip signal from an electrical discharge from the turbine shaft. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-9053.
Failure to fabricate the turbine shaft grounding brush in accordance with vendor instructions is a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected 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. Specifically, the improperly fabricated grounding brush resulted in a turbine trip and subsequent reactor scram. The inspectors reviewed the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affecting mitigating equipment, and did not increase the likelihood of a fire or flood. The apparent cause of the performance deficiency was the failure in 2004 to appropriately perform a post maintenance test for the turbine shaft grounding brush modification. Therefore the inspectors did not identify a cross-cutting aspect because the performance deficiency is not reflective of the licensees current performance.
Inspection Report# : 2012002 (pdf)
Mitigating Systems Significance:        Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of the Control Building Chilled Water System The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the failure to maintain temperature control of the safety-related battery rooms. An engineering evaluation to change a procedure to allow gagging open of the control building heating and ventilation system control temperature valves failed to consider the appropriate environmental temperature limits for the rooms. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-07353.
The failure to maintain temperature control of the safety-related battery rooms was a performance deficiency. This performance deficiency is more-than-minor and is therefore a finding because it is associated with the design control attribute of the mitigating systems cornerstone and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, during a loss of offsite power with low seasonal temperatures, the gagged-open temperature control valve would reduce the battery rooms temperatures below their environmental design temperature and adversely affect the capacity of the safety-related batteries. In accordance with NRC Inspection Manual Chapter 0609, , Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, Section A.1, this finding screened as very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality. The engineering evaluation that changed the proper battery room controls was performed in 1997. Therefore, the finding did not have a cross-cutting aspect because the failed review is not indicative of current licensee performance.
Inspection Report# : 2012005 (pdf)
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4Q/2012 Inspection Findings - River Bend 1 Significance:        Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Lubrication of the Standby Liquid Control Pump Motor Bearings The inspectors identified a non-cited violation of Technical Specification 5.4.1.a for not establishing appropriate lubrication procedures for the standby liquid control pump motor bearings. Specifically, the station incorrectly used the Electrical Power Research Institute (EPRI) guidance for maintenance procedure by adding twice the amount of grease required. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-05573.
The failure to establish appropriate lubrication procedures is a performance deficiency. This performance deficiency is more-than-minor and is therefore a finding because if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, if the work instructions were not corrected, future work activities that grease the motor bearings in accordance with those work orders would over-grease the bearings, which may result in common-cause failures of standby motors. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, Section A.1, this finding screened as very low safety significance (Green). Specifically, the finding is a deficiency that affected the qualification of the standby liquid control pump motors; however, the systems maintained their operability. Because the most significant causal factor of the performance deficiency was station personnel and management failing to fully evaluate the previously identified inadequate lubrication of motors, this finding has a problem identification and resolution cross-cutting aspect associated with the corrective action program component [P.1(c)].
Inspection Report# : 2012005 (pdf)
Significance:        Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure for Lifting Leads Results in Inoperability of Standby Service Water Fan The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a due to a failure to follow work order instructions. Specifically, station personnel failed to follow the requirements of Procedure GMP-0042, Lifted Leads and Jumpers, Revision 13 when removing and reinstalling a time-delay relay for a standby service water cooling fan. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-06325.
The failure to follow work order instructions is a performance deficiency. This performance deficiency is more-than-minor because it is associated with the equipment reliability attribute of the mitigating systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure the correct wiring to the standby service water fan time-delay relay resulted in the inability of the fan to be started locally, which is required for remote shutdown of the plant. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, Section A, question 3, this finding required a detailed risk evaluation because the finding represented an actual loss of function of at least a single train for greater than the technical specification allowed outage time. The risk of the condition was evaluated by a senior reactor analyst. The sequence that would result in a risk increase is control room abandonment with concurrent maintenance being performed on the alternate bank of 5 fans. This would leave only 4 functional fans in one division of standby service water, whereas 5 fans are needed per design to meet the safety function.
The frequency of control room abandonment is approximately 5E-5/yr and the frequency of maintenance performed on one bank of standby service water fans is approximately 1E-2. Therefore, the frequency of a scenario where the failure of one fan to operate from the alternate shutdown panel would cause a measurable effect on risk is approximately 5E-7/yr. The other division of standby service water fans was unaffected by this condition.
Accordingly, the significance of the performance deficiency was determined to be very low (Green). This finding has a human performance cross-cutting aspect associated with the work practices component in that the electricians failed Page 6 of 15
 
4Q/2012 Inspection Findings - River Bend 1 to use adequate human error prevention techniques [H.4(a)].
Inspection Report# : 2012005 (pdf)
Significance:        Oct 11, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct Spurious Isolations of Reactor Core Isolation Cooling System The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality.
Specifically, the licensee failed to identify and correct an inadequate design of the reactor core isolation cooling (RCIC) system that resulted in spurious system isolations during main turbine trips. In response, the licensee installed a time delay into the circuit that had tripped the RCIC steam supply before the RCIC received a start signal. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-03439.
The performance deficiency was more than minor because it was associated with the design control 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, in that the repeated spurious isolations adversely affected the RCIC system reliability. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, this finding screened to a detailed risk evaluation which determined that the finding was of very low safety significance (Green). This finding does not have a cross-cutting aspect because the apparent cause of this finding was the licensees decision in 2008 to not add a time delay to the high differential pressure trip, and the NRC does not consider that cause to be representative of current licensee performance (Section 4OA5.2.a).
Inspection Report# : 2012010 (pdf)
Significance:        Oct 11, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Declare Reactor Core Isolation Cooling System Inoperable The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to declare the RCIC system inoperable when the system was unreliable for an automatic start following a main turbine trip. The licensee addressed the underlying safety concern by installing a time delay into the circuit that had tripped the RCIC steam supply before RCIC received a start signal. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-06015.
The performance deficiency was more than minor because it affected 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. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, this finding screened to a detailed risk evaluation which determined that the finding was of very low safety significance (Green). Because the most significant causal factor of the performance deficiency was that the organization had used the absence of information to determine RCIC operability, this finding has a cross-cutting aspect in the human performance area associated with the decision-making component, because the licensee had failed to demonstrate that the proposed action was safe in order to proceed rather than a requirement to demonstrate that it was unsafe in order to disapprove the action [H.1(b)] (Section 4OA5.2.b).
Inspection Report# : 2012010 (pdf)
Significance:        Oct 11, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Prrevent Conflicts of Duty for Fire Brigade Members Page 7 of 15
 
4Q/2012 Inspection Findings - River Bend 1 The inspectors reviewed a self-revealing, non-cited violation of License Condition 2.C.(10) because the licensee failed to prevent conflict of duties for fire brigade members, which affected the timely response to fires. In response, the control room initiated a night order to ensure that when a fire brigade member is called for fitness-for-duty testing, the staff will either designate a relief fire brigade member or arrange a deferral of the fitness-for-duty testing. The licensee plans to address long-term corrective actions through appropriate procedure changes at the fleet level. The licensee entered the finding into the corrective action program as Condition Report CR RBS-2012-03817.
The performance deficiency was more than minor because it was associated with the protection against external events 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. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, this finding screened to very low safety significance (Green) because the affected fire brigade member was unavailable for less than two hours. Because the most significant causal factor of the performance deficiency was that the licensee failed to ensure that conflicts between the fitness-for-duty and fire brigade procedures had been properly resolved prior to implementation, this finding has a human performance cross cutting aspect associated with resources because the licensee did not ensure that procedures were complete and accurate to assure nuclear safety [H.2(c)] (Section 4OA5.8).
Inspection Report# : 2012010 (pdf)
Significance:        Sep 28, 2012 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions to Ensure Reliability of the 480 VAC Molded Case Circuit Breakers and Unitized Motor Starters The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct within a reasonable period conditions adverse to quality associated with testing safety-related molded-case circuit breaker and unitized motor starter circuit breakers. The licensees immediate corrective actions included increasing the rate of breaker preventive maintenance and testing to reduce the long-standing risk-significant breaker backlog. The station documented the finding in Condition Report CR-RBS-2012-06364.
The performance deficiency was more-than-minor and is therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the safety-related molded-cased circuit breakers to respond to initiating events to prevent undesirable consequences. Specifically, failures of the affected breakers represent an increase in risk to safe plant operations, because to isolate a fault caused by a defective 480VAC breaker, the upstream feeder breaker would trip, thus causing a loss of power to additional safety-related components. Using Inspection Manual Chapter 0609, Appendix A, the finding is associated with the loss of mitigation equipment (Service Water pumps A and C), and so screened to a detailed risk evaluation. That evaluation determined that the incremental conditional core damage probability (ICCDP) was 2.1E-8 for a fire in one of the standby cooling tower electrical rooms, resulting in a loss of one train of service water pumps (A and C, or B and D), as a consequence of the failure of the proximate 480 VAC breaker to open. The risk was low because normal service water would be unaffected by the fire, and it would be unlikely that offsite power would be lost concurrently. The fire could also affect control room ventilation, but the analyst qualitatively concluded that this would not add more than negligibly to the overall risk. Consequently, the finding has very low safety significance (Green). The inspectors determined that the apparent cause of the finding was a combination of two factors related to resources: station management did not ensure that each work group completed its actions to support timely resolution, and personnel vacancies from key positions hampered completion of the breaker testing program. The inspectors therefore determined the finding had a cross-cutting aspect in the human performance area associated with the resources component because station management did not ensure personnel resources were available to minimize long-standing equipment issues. [H.2(a)].
Inspection Report# : 2012004 (pdf)
Significance:        Sep 28, 2012 Identified By: NRC Page 8 of 15
 
4Q/2012 Inspection Findings - River Bend 1 Item Type: NCV NonCited Violation Failure to Appropriately Tune the Reactor Core Isolation Cooling Turbine Speed Controller The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with inadequate instructions for tuning the reactor core isolation cooling (RCIC) terry turbine speed governor. The licensees immediate corrective actions included revising the maintenance procedure and recalibrating the RCIC turbine speed controller. The station documented the finding in Condition Reports CR-RBS-2012-01750 and CR-RBS-2012-01904.
This performance deficiency is more-than-minor and is therefore finding because it is associated with the equipment performance 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, during operation, this performance deficiency resulted in improper tuning of the turbine speed control system, which caused the turbine exhaust check valve to repeatedly slam against its open and shut valve stops and abnormally large turbine governor valve oscillations. Because the licensee had not tuned the turbine speed control system to run at a steady speed, the licensee removed RCIC from service to properly calibrate the control system, thereby adversely affecting RCIC availability. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification; did not represent a loss of system and/or function, did not represent either an actual loss of function of at least a single train for greater than its Technical Specification Allowed Outage Time, or two separate safety systems out-of-service for greater than its Technical Specification Allowed Outage Time; and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined the apparent cause of this finding was the licensees failure to incorporate industry and vendor operating experience into the work instructions on February 12, 2011, to correct RCIC governor valve oscillations. Therefore, this finding has a cross-cutting aspect in the area of problem identification and resolution associated with the operating experience component because the licensee did not implement and institutionalize industry knowledge, including vendor recommendations, to support plant safety [P.2 (b)].
Inspection Report# : 2012004 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: NCV NonCited Violation High Pressure Core Spray Diesel Generator Bearing Lubrication Deficiencies The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failing to correct a condition adverse to quality for lubricating the high pressure core spray diesel generator bearings.
The station documented the finding in Condition Report CR-RBS-2012-02666.
This performance deficiency was more than minor and was a finding because, if left uncorrected, inadequate lubrication work instruction could cause bearing failure due to inadequate lubrication or generator winding failure due to grease intrusion into the electrical windings in the generator. The significance of this finding was evaluated using a Phase 1 significance determination process screening and was determined to be of 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 seismic, flooding, or severe weather initiating events. The apparent reason the initial condition report was closed without correcting the work instruction to lubricate the high pressure core spray diesel generator bearings was that personnel who prepared and approved the operability evaluation were focused on proving operability not correcting a condition adverse to quality. Their focus was specific to the components ability to perform its function and not on completely identifying the issue in the corrective action program. Therefore, the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the station did not identify this issue completely, accurately, and in a timely manner commensurate with its safety significance [P.1(a)].
Inspection Report# : 2012003 (pdf)
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4Q/2012 Inspection Findings - River Bend 1 Significance:      Jun 29, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Specify Manual Actions for Safety Relief Valve Operations During a Station Blackout Event The inspectors identified a non-cited violation of 10 CFR 50.63, Loss of All Alternating Current, paragraph (a) (2),
which states, in part, The reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. The capability for coping with a station blackout of specified duration shall be determined by an appropriate coping analysis. Licensees are expected to have the baseline assumptions, analyses, and related information used in their coping evaluations available for NRC review. Specifically, from November 1985 to May 17, 2012, the licensee failed to specify actions while ac power is unavailable to ensure that safety relief valves provided sufficient capacity and capability to ensure appropriate containment integrity is maintained during a station blackout event. This violation has been entered into the corrective action program as Condition Report CR-RBS-2012-03376.
The inspectors determined that failure to specify actions for safety relief valve operation in procedures in accordance with NUMARC-8700 was a performance deficiency. The finding was more than minor because it adversely affected the procedure quality attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to respond to undesirable consequences. Specifically, the station blackout coping procedures did not specify actions that would ensure the heat capacity temperature limit for the suppression pool would not be exceeded during the station blackout coping period. Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the inspectors determined that the Mitigating Systems Cornerstone was affected because the finding could cause degradation of core decay heat removal. Using Table 4a from the Phase 1 worksheet, the inspectors determined that the finding represents a loss of safety function; therefore, a Phase 2 analysis was necessary. However, the inspectors determined that a Phase 2 analysis was not sufficient to assess significance because of the complexity of the finding. Therefore, a Phase 3 analysis was necessary. The result of the Phase 3 analysis determined that the change in core-damage-frequency (?CDF) for the performance deficiency was 2.4E-7 or very low safety significance (Green). The senior reactor analyst determined that the change in large-early-release-frequency (?LERF) was 4.8E-8 or very low safety significance (Green). No cross-cutting aspect was identified because the most significant contributor was not indicative of current licensee performance (Section 4OA5).
Inspection Report# : 2012003 (pdf)
Significance:      Mar 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Appropriately Assess and Manage Risk for Internal Flooding Events 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, due to the failure of work control and operations personnel to adequately assess the increase in risk associated with internal flooding events. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2012-00641.
The failure of work control and operations personnel to adequately assess the risk associated with internal flooding is a performance deficiency. The performance deficiency resulted in the overall elevated plant risk placing the plant into the higher licensee-established risk category (Green to Yellow). The performance deficiency 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. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowcharts 1 and 2, the finding was determined to have very low safety significance (Green) because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability deficit was less than 1E-7. The inspectors determined that the apparent cause of the finding was that station personnel routinely failed to review the qualitative risk checklist required by the stations risk management procedure. Therefore, this finding has a cross-cutting aspect in the human performance area associated with the work practice component because the licensee did not define and effectively communicate expectations regarding procedural compliance. [H.4(b)].
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4Q/2012 Inspection Findings - River Bend 1 Inspection Report# : 2012002 (pdf)
Significance:      Mar 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify Assumptions used in Standby Equipment Room Temperature Analysis The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because, prior to February 7, 2012, the licensee did not verify that assumptions used in confirming that the safety-related battery inverter rooms would remain below their design basis temperature limits during a design basis event agreed with the as-built condition of the plant. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2012-01046.
The inspectors determined that the failure to verify that design documents match the actual configuration of the plant is a performance deficiency. The finding was more than minor because it adversely affects the Mitigating Systems Cornerstone objective of equipment performance to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee had not verified assumptions that ensure the standby switchgear room air conditioning system would reliably maintain the standby equipment rooms below the design temperature limits. Using Inspection Manual Chapter 0609, Attachment 4, "Initial Screening and Characterization of Findings," the finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function, nor actual loss of safety function of a single train, and it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that this issue has a cross-cutting aspect in the area of human performance decision-making regarding nonconservative assumptions. When the licensee conducted the flow balance test, they assumed that measuring air inflow alone was sufficient, but did not check that the doors gaps were allowing a sufficient amount of warm air to exit standby equipment rooms and be circulated back to the general areas [H.1(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Appropriately Set Reactor Core Isolation Cooling Flow Controller High Output Limit The inspectors identified a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Procedures, when the reactor core isolation cooling turbine tripped on mechanical over speed. Troubleshooting determined the cause was an improperly tuned flow controller. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2012-01188 and CR-RBS-2012-01262.
The failure to provide specific flow controller tuning instructions for the reactor core isolation cooling turbine flow controller was a performance deficiency. The finding was more than minor in accordance with Appendix B, "Issue Screening," of Inspection Manual Chapter IMC 0612, "Power Reactor Inspection Reports," because the finding was associated with the equipment performance 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, improper tuning of the reactor core isolation cooling controller impacted operability and availability of the reactor core isolation cooling system. The inspectors performed a Phase 1 significance determination process review of this finding per Inspection Manual Chapter IMC 0609, Attachment 4, "Initial Screening and Characterization of Findings." In accordance with Table 4a, "Characterization Worksheet for IE, MS, and BI Cornerstones," the finding represented a loss of system safety function. Therefore, a Region IV senior reactor analyst used Inspection Manual Chapter IMC 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," to review the finding using the Standardized Plant Analysis Risk (SPAR) model for River Bend Station. The Phase 3 analysis determined the Delta-CDF was 4.68E-7/yr. For a 7-month exposure, the incremental conditional core damage probability is 2.73E-7. The majority of the risk came from sequences involving a loss of feedwater (48 percent) and a loss of offsite power (33 percent). Consequently, the analyst determined that the risk associated with the performance deficiency was very low (green). The inspectors determined the apparent cause of this finding was the failure to perform a post maintenance test to identify that the high output limit was not properly set by the maintenance work instruction. Therefore, this finding has cross-cutting Page 11 of 15
 
4Q/2012 Inspection Findings - River Bend 1 aspect in the area of human performance associated with the resources component due to less than adequate work package testing instruction. [H.2(c)].
Inspection Report# : 2012002 (pdf)
Barrier Integrity Significance:        Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Effective Corrective Actions for Defects in MasterPact Breakers The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to promptly correct a condition adverse to quality. Specifically, station personnel failed to implement repairs to the mechanism-operated contact linkages for safety-related breakers, ultimately resulting in the failure of standby gas treatment filtration train 1B to start on demand. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-005894.
The failure to correct a condition adverse to quality is a performance deficiency. This performance deficiency is more-than-minor because it is associated with the systems, structures, and components and barrier performance attributes of the barrier integrity cornerstone and adversely affected the 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. Specifically, the standby gas treatment exhaust filter train failed to start during a surveillance test because of a nonconforming mechanical linkage in the feeder breaker resulting in unavailability for standby gas train 1B. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 3, Section C, question 1, the finding screened as very low safety significance (Green), because the finding represented only a degradation of the radiological barrier function provided by the standby gas treatment system. No cross-cutting aspect was assigned to this finding because the NRC concluded the finding did not reflect current licensee performance.
Inspection Report# : 2012005 (pdf)
Significance:        Sep 28, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Improper Hydrogen Igniter Breaker Trip Coil Setting The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the design bases for the power supply for the hydrogen igniter system into procedures used to set the associated power system supply breaker trip coil. The licensees immediate corrective actions included evaluating the proper trip coil setting and adjusting the trip coil accordingly. The station documented the finding in Condition Report CR-RBS-2012-02623.
This performance deficiency is more-than-minor and is therefore a finding because it is associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, this performance deficiency resulted in an incorrect trip coil setting, which decreased the reliability of the hydrogen igniters, which burn hydrogen in a controlled manner to prevent containment damage. Using Inspection Manual Chapter 0609.04, Initial Characterization of Findings, the finding required a significance evaluation per Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, because the unavailable Division 1 hydrogen igniters represented a degraded condition affecting containment barrier integrity that can potentially increase large early release frequency (LERF) without affecting the core damage frequency (CDF). Inspectors determined that this was a type B finding. Using section 6.0, the inspectors determined that the finding was of very low safety significance (Green) because the hydrogen igniters are arranged in two independent divisions such that each containment region has two igniters, one Page 12 of 15
 
4Q/2012 Inspection Findings - River Bend 1 from each division, controlled and powered redundantly so that ignition would occur in each region even if one division failed to energize. The inspectors determined that the apparent cause of this finding was that in response to earlier failures of the trip coil, the licensee had not investigated the problem thoroughly enough to identify and correct this performance deficiency. However, because the earlier failures had all occurred more than seven years ago, the inspectors determined that this cause did not reflect present licensee performance, so the inspectors did not assign a cross-cutting aspect to it.
Inspection Report# : 2012004 (pdf)
Significance:        Mar 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Maintenance Instructions used for Suppression Pool Cooling Isolation Valve Maintenance The inspectors identified a Green, self-revealing non-cited violation of Technical Specification 5.4.1.a, Procedures, for inadequate maintenance procedures to properly assemble containment isolation valves on the suppression pool cooling system. This resulted in a failure of the suppression pool cooling systems outboard containment isolation valve marriage coupling that ensures the valve stem is connected to the valve actuator. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2011-09171.
The failure to establish adequate work instructions to assemble the suppression pool cleanup system isolation valves is a performance deficiency. The inspectors determined that the finding was more than minor because it is associated with the Barrier Integrity Cornerstone attribute of Systems, Structures, and Components and Barrier Performance, and affected the cornerstone objective of providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations. Using the Phase 1 SDP worksheet for the barrier integrity cornerstone, the inspectors answered no to all four screening questions under the containment barrier column. Specifically, the affected penetration did not represent an actual open pathway in the physical integrity of reactor containment due to an operable and functionally redundant containment isolation valve in the suppression pool cooling piping penetration. The apparent cause of the finding was the failure of the planning department to recognize and develop design documentation to identify the set screw size and starting material necessary to determine the appropriate set screw torque for work affecting safety related equipment. The inspectors determined the finding had a cross cutting aspect in the human performance, area associated with the resources component because of the lack of complete accurate and up to date design documentation associated with the work package development. [H.2(c)].
Inspection Report# : 2012002 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Mar 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform a Radiation Survey Inspectors reviewed a self-revealing non-cited violation of 10 CFR 20.1501(a) for the failure to perform a radiation survey. A survey was not completed after two contaminated valves were transferred from the 98-foot elevation of the main steam tunnel to the radwaste area. During shift turnovers, workers responsible for transferring the valves did not understand that they needed to remove two buckets, and perform a survey after completing the valve transfer.
Consequently, a bucket with highly contaminated water and residual was left in the tunnel causing radiation levels as high as 300 millirem per hour. This resulted in an unposted high radiation area. The licensee entered the issue into the Page 13 of 15
 
4Q/2012 Inspection Findings - River Bend 1 corrective actions program as Condition Report CR-RBS-2011-01552.
The failure to perform a radiation survey to evaluate the radiological conditions is a performance deficiency. The finding is more than minor because it negatively impacted the Occupational Radiation Safety cornerstones attribute of program and process, in that the lack of a post-work survey did not ensure exposure control for workers. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of 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 the ability to assess dose was not compromised. The finding has a Human Performance cross-cutting component associated with the aspect of work practices because expectations regarding procedural compliance for post-job radiation surveys were ineffective [H.4(b)].
Inspection Report# : 2012002 (pdf)
Significance:      Mar 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Control Access to a High Radiation Area Inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.7.1(c), resulting from the licensees failure to control access to a high radiation area. Specifically, a carpenter entered a high radiation area in the main steam tunnel near valve V112 without proper authorization before a health physics technician completed radiation surveys and received an unexpected alarming dosimeter reading of 110 millirem per hour. The carpenter had not been briefed that dose rates in the area measured 140 millirem per hour. He had been instructed not to perform any work before the health physics technician surveyed the area, but River Bend did not make it clear enough that he was to follow all health physics instructions. The licensee entered the issue into the corrective actions program as Condition Report CR-RBS-2011-01426 and the worker was counseled.
The failure to control access to a high radiation area was a performance deficiency. The finding was more than minor because it was associated with the occupational radiation safety attribute of exposure control and affected the cornerstone objective in that not controlling a high radiation area could increase personal exposure. In addition, this type of issue is addressed in Example 6.h of IMC 0612, Appendix E, Examples of Minor Issues. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The finding has a Human Performance cross-cutting component associated with the aspect of work practices because expectations regarding supervisory and management oversight of work activities, including contractors to ensure that safety is supported were not met [H.4(c)].
Inspection Report# : 2012002 (pdf)
Public Radiation Safety Significance:      Mar 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Digital Radiation Monitoring System The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of the digital radiation monitoring system. Specifically, the maintenance rule expert panel performed an inadequate analysis after the digital radiation monitoring system exceeded the condition monitoring criteria by failing to follow the procedural requirements of EN-DC-206 to have cause evaluations for system failures so that maintenance preventability could be properly evaluated. This issue has been entered into the licensees corrective action program as Condition Reports CR-RBS-2011-00485.
The inspectors determined that the failure to adequately monitor the performance of the digital radiation monitoring Page 14 of 15
 
4Q/2012 Inspection Findings - River Bend 1 system is a performance deficiency. The inspectors reviewed Inspection Manual Chapter (IMC) 0612 and determined that the finding is more than minor because the finding is associated with the plant facilities/equipment and instrumentation attribute (reliability of process radiation monitors) of the radiation safety cornerstone (public radiation safety) and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian use. The finding was assessed using the IMC 0609, Appendix D, Public Radiation SDP, and because there was no failure to implement the effluent program, the finding was determined to be of very low safety significance (Green). The inspectors reviewed the apparent cause of this finding and found that the oversight of the maintenance rule program was adversely affected by personnel changes and lack of effective turnover. Therefore, the finding has a cross-cutting aspect in the human performance area and resources component because the licensee failed to ensure that maintenance rule program personnel were trained and sufficiently qualified to perform their duties in an effective manner [H.2(b)].
Inspection Report# : 2012002 (pdf)
Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance:        Oct 11, 2012 Identified By: NRC Item Type: NCV NonCited Violation Loss of Onsite Safety Review Committee Independence The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, for several examples of failures to follow Procedure EN OM 119, Onsite Safety Review Committee, Revision 8, which indicated that the onsite safety review committee failed to accomplish an independent review of station activities in accordance with the procedure. In response to this finding, the licensee developed a process to document the committee findings and reinforced roles and responsibilities for committee conduct, and committee members reviewed the implementing procedure. The licensee entered this finding into the corrective action program as Condition Report CR-RBS-2012-03739.
The multiple failures to follow the onsite safety review committee implementing procedure were performance deficiencies that were more-than-minor because failure to correct these performance deficiencies could compromise the nuclear safety oversight function of the committee, which could result in inappropriate decision-making on activities important to nuclear safety. In accordance with NRC Inspection Manual Chapter 0609 Appendix M, Significance Determination Process Using Qualitative Criteria, the finding was of very low safety significance because the performance deficiency did not result in any risk-significant issues. Because the most significant causal factor of the performance deficiency was the licensees failure to properly define, communicate and implement the roles for decision-making that affected nuclear safety, this finding has a human performance cross-cutting aspect associated with decision-making because the licensee failed to adequately communicate the authority and roles of the onsite safety review committee to the members [H.1(a)] (Section 4OA5.7).
Inspection Report# : 2012010 (pdf)
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1Q/2013 Inspection Findings - River Bend 1 River Bend 1 1Q/2013 Plant Inspection Findings Initiating Events Significance:        Oct 11, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish an Adequate Controlling Procedure for Stroking Safety Relief Valves at Low Power The inspectors identified a Green non-cited violation of Technical Specification 5.4.1.a for the failure to develop adequate controls for low-power stroking of safety relief valves. In response to this finding, the licensee trained senior reactor operators on the lessons learned from the finding. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-03816.
The performance deficiency was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Section B, this finding screened to very low safety significance because it was a transient initiator that did not result in a reactor trip and loss of mitigation equipment. Because the most significant causal factor of the performance deficiency was that the licensee had made an inappropriate assumption that the abnormal operating procedure was a satisfactory controlling document, this finding has a human performance cross cutting aspect associated with the decision making component, in that the licensee failed to use conservative assumptions in decision-making [H.1(b)] (Section 4OA5.3).
Inspection Report# : 2012010 (pdf)
Significance:        Oct 11, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Effective Corrective Actions for Lockout Relay Failures The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality. Specifically, after a lockout relay mechanically bound in 2011, causing a fire, the licensee failed to identify and correct other susceptible relays. In response, the licensee tested other susceptible relays and replaced those that failed the test. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-05894.
This performance deficiency was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Section B, this finding screened to a detailed risk evaluation because it had caused a reactor trip and the loss of mitigation equipment such as loss of main feedwater and normal service water. The detailed risk evaluation included a quantitative bounding analysis and a qualitative evaluation in accordance with NRC Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, to determine Page 1 of 17
 
1Q/2013 Inspection Findings - River Bend 1 that this finding was of very low safety significance (Green). Because the most significant causal factor of the performance deficiency was that the licensee had failed to recognize the potential risk to the plant when performing the evaluations for the failed lockout relays, this finding has a human performance cross-cutting aspect associated with the work control component in that licensee did not plan and coordinate work activities by incorporating risk insights, consistent with nuclear safety [H.3(a)] (Section 4OA5.4).
Inspection Report# : 2012010 (pdf)
Significance:        Oct 11, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Test Lockout Relays in Accordance with Vendor Testing Practices The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, for the licensees failure to establish adequate preventative maintenance instructions for lockout relays in accordance with vendor recommendations for electrical testing. In response, the licensee incorporated vendor recommendations into the instructions for testing lockout relays. The licensee entered the finding into the corrective action program as Condition Report CR RBS-2011-02209.
The performance deficiency was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, in that it resulted in a fire. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Section B, this finding screened to very low safety significance (Green) because it was a transient initiator that did not result in a reactor trip or loss of mitigation equipment. The finding did not have a cross-cutting aspect because the performance deficiency was not representative of current plant performance (Section 4OA5.5).
Inspection Report# : 2012010 (pdf)
Significance:        Oct 11, 2012 Identified By: NRC Item Type: FIN Finding Failure to Establish An Adequate Cable Reliability Program The inspectors reviewed a self-revealing finding for the licensees failure to establish an effective cable reliability program, in that the licensee failed to distinguish between wetted and dry splices. In response, the licensee tested the high-risk-ranked cables, and replaced those that failed the test. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-03440.
The performance deficiency was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, in that it resulted in a reactor scram. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 1, Section B, this finding screened to very low safety significance (Green) because it was a transient initiator that did not result in both a reactor trip and loss of mitigation equipment. Because the most significant causal factor of the performance deficiency was that the licensee failed to implement and institutionalize operating experience related to wetted splices, this finding has a problem identification and resolution cross cutting aspect associated with operating experience in that the licensee did not implement and institutionalize operating experience through changes to station processes and procedures to support plant safety [P.2 (b)] (Section 4OA5.6).
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1Q/2013 Inspection Findings - River Bend 1 Inspection Report# : 2012010 (pdf)
Significance:        Sep 28, 2012 Identified By: NRC Item Type: FIN Finding Inadequate Verification of Leading Edge Flow Meter Functionality The inspectors identified a finding for the licensees failure to calibrate the feed water Leading Edge Flow Meter (LEFM) CheckPlus System following maintenance activities. This resulted in an error in reactor feed water flow rate data used to calculate reactor core thermal power. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-06274.
This performance deficiency is more-than-minor and is therefore a finding because it was associated with the procedure quality attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The performance deficiency challenged the initiating events cornerstone objective by allowing the licensee to operate the plant outside of the prescribed analyzed uncertainty value, used in determining maximum core thermal power. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that this finding has very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the apparent cause of this finding was that when the licensee had changed the flow meter maintenance work scope that required transducer replacement, they had not included the vendor verification requirement in the revised work order. Therefore, this finding has a cross-cutting aspect in the Human Performance area of Work Control because the licensee had failed to appropriately coordinate the impact of changes to the work scope or activity on the plant. [H.3(b)].
Inspection Report# : 2012004 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: FIN Finding Failure to Follow Procedure to Protect Sensitive Plant Areas The inspectors identified a finding for failure to follow Operating System Procedure OSP-0048, "Switchyard, Transformer Yard, and Sensitive Equipment Controls." Specifically, the licensee failed to appropriately consider the plant impact when planning and approving work in the main transformer yard and switchyard potentially introducing unacceptable risk to plant operations contrary to OSP-0048 administrative controls. This issue was entered into the licensees corrective action program as Condition Reports CR-RBS-2012-02479, CR-RBS-2012-02821, and CR-RBS-2012-04129.
The finding was more than minor in accordance with Appendix B, "Issue Screening," of Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," because the finding was associated with the protection against external events attribute of the Initiating Events cornerstone and affected 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. Specifically, the routine failure to integrate switchyard and transformer yard work into the River Bend work process increased the likelihood that unintended, uncoordinated maintenance and test activities could reduce the diversity of electrical power and cause inadvertent reductions in nuclear plant defense-in-depth. The inspectors performed a Phase 1 significance determination process review of this finding per Inspection Manual Chapter 0609, , "Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) since the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, nor did it contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, and the finding did not increase the likelihood of a fire or Page 3 of 17
 
1Q/2013 Inspection Findings - River Bend 1 internal or external flooding. The inspectors determined the apparent cause of this finding was a lack of management oversight of station work activities. Therefore, this finding has a cross-cutting aspect in the area of human performance associated with the work practices component because station management failed to provide proper oversight of the process to protect sensitive areas of the plant [H.4(c)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: FIN Finding Failure to Implement Severe Weather Operations Procedure The inspectors identified a finding that involved failure to implement a procedure to protect the plant during adverse weather conditions. Specifically, appropriate equipment walkdowns and corrective actions were not performed to protect equipment important to safety from severe weather risks in a timely manner. The concerns were documented in Condition Report CR-RBS-2012-02387.
The finding was determined to be of very low safety significance (Green) since the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator, nor did it contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, and the finding did not increase the likelihood of a fire or internal or external flooding. The inspectors determined the apparent cause of this finding was operations expectation that excellent housekeeping nominally exists in the switchyard and transformer yard. Therefore, there was no need to dispatch personnel to verify housekeeping because that action would risk personnel safety. The status of an unsecured ladder in the transformer yard is evidence that up to date information is essential to confirm whether housekeeping is satisfactory. Therefore, the finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the station did not demonstrate that nuclear safety was an overriding priority because it failed to implement the roles and authorities in their severe weather operations procedure [H.1(a)].
Inspection Report# : 2012003 (pdf)
Significance:        Jun 29, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Properly Assemble Turbine Control Valve Push Rod-Spring Housing Coupling The inspectors reviewed a self-revealing finding associated with main turbine control valve number 3 unexpectedly closing. In response, operators reduced reactor power to 90 percent. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-02773.
The finding was more than minor because it was associated with the Initiating Events cornerstone attribute of design control and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability by resulting in a plant downpower and subsequent planned outage for repair activities. The inspectors reviewed the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. Based on the Phase 1 screening of the finding, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect loss of coolant accident initiators, did not contribute to increasing the likelihood of both an initiating event and affecting mitigating equipment, and did not increase the likelihood of a fire or flood. The inspectors did not identify a cross cutting aspect because the performance deficiency is not indicative of the licensees current performance.
Inspection Report# : 2012003 (pdf)
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1Q/2013 Inspection Findings - River Bend 1 Mitigating Systems Significance:        Mar 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor the Performance of the Floor and Equipment Drains System The inspectors identified a non-cited violation of 10 CFR 50.65(a)(1) associated with the licensees failure monitor the floor and equipment drains system against licensee-established goals. The licensee failed to properly classify two maintenance preventable functional failures for this system, and as a result, inappropriately left the system in maintenance rule a(2) status. In response, the licensee properly classified the subject failures and classified the affected system into maintenance rule (a)(1) status. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2013-00295.
The failure to adequately monitor the performance of the floor and equipment drains system is a performance deficiency. The performance deficiency was more-than-minor and was therefore a finding because if left uncorrected, the failure to adequately monitor the performance of the floor and equipment drains system could lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At-Power, the inspectors determined that the finding is of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety significance in accordance with the licensees maintenance rule program. No cross-cutting aspect was assigned because the finding does not represent current performance.
Inspection Report# : 2013002 (pdf)
Significance:        Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Identify and Correct a Condition Adverse to Quality The inspector documented a self-revealing finding associated with the licensees failure to follow the requirements of Station Procedure EN-LI-102, Corrective Action Process, and promptly identify and correct a condition adverse to quality. Specifically, on August 4, 2010, and again on February 14, 2011, station personnel found where the B reactor feedwater pumps auxiliary oil system pressure regulator set point had drifted high out of tolerance, but did not initiate condition reports for this condition adverse to quality. The licensee entered this issue into their corrective action program as Condition Reports CR-RBS-2011-09141 and CR-RBS-2012-07249.
The failure to follow the requirements of Station Procedure EN-LI-102 and identify and correct a condition adverse to quality was a performance deficiency. The performance deficiency was more than minor, and is therefore a finding because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At-Power, the inspector determined that the finding is of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than its technical specification allowed outage time; and (4) did not represent an actual loss Page 5 of 17
 
1Q/2013 Inspection Findings - River Bend 1 of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. This finding had a cross-cutting aspect in the area of human performance associated with the work practices component, in that, the licensee failed to define and effectively communicates expectations regarding procedural compliance and personnel follow procedures. Specifically, station personnel failed to follow procedure.
Inspection Report# : 2012012 (pdf)
Significance:      Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify and Correct a Condition Adverse to Quality The inspector identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, following a turbine trip/load reject and subsequent reactor scram, reactor vessel level rose to the point of receiving a high level isolation signal (Level 8), and the licensee failed to identify this as an unexpected condition. The licensee entered this issue into the corrective action program as Condition Report CR-RBS-2012-07250.
The failure to promptly identify and correct a condition adverse to quality was a performance deficiency. The performance deficiency is more than minor, and is therefore a finding because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, the inspector determined that the finding is of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action component, because the licensee failed to periodically trend and assess information from the corrective action program and other assessments in the aggregate to identify programmatic and common cause problems.
Inspection Report# : 2012012 (pdf)
Significance:      Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of the Feedwater Control System The inspector identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to ensure that design requirements were correctly translated into installed plant equipment. Specifically, the licensee failed to appropriately translate the feedwater control systems design of maintaining full feedwater capacity following a turbine trip with load rejection by avoiding loss of feedwater due to a high level isolation (Level 8) using the level set point modification module. The licensee entered this issue into the corrective action program as Condition Report CR-RBS-2012-02249 and CR-RBS-2012-07254.
The failure to ensure that design requirements were correctly translated into installed plant equipment was a performance deficiency. This performance deficiency is more than minor, and is therefore a finding, because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that Page 6 of 17
 
1Q/2013 Inspection Findings - River Bend 1 respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At-Power, the inspector determined that the finding is of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. This finding did not have a cross cutting aspect because the most significant contributor did not reflect current licensee performance.
Inspection Report# : 2012012 (pdf)
Significance:        Dec 31, 2012 Identified By: NRC Item Type: FIN Finding Failure to Correct an Identified Condition Adverse to Quality The inspector identified a finding associated with the licensees failure to follow the requirements of Station Procedure EN-LI-102, Corrective Action Process, and correct a condition adverse to quality. Specifically, the licensee identified that both inadequate guidance and oversight of a supplemental worker as a cause for the inadequate crimp on the B reactor feedwater pump, however the corrective actions taken only addressed the oversight of supplemental workers, and no actions were taken to address the insufficient guidance provided by the station work order. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2012-07253.
The failure to follow the requirements of Station Procedure EN-LI-102 and correct a condition adverse to quality was a performance deficiency. The performance deficiency is more than minor, and is therefore a finding, because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At-Power, the inspector determined that the finding is of very low safety significance (Green). because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action component, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes.
Inspection Report# : 2012012 (pdf)
Significance:        Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct the Maintenance Organizations Inadequate Procedure Use and Adherence The inspector identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, during a root cause evaluation associated with a lockout relay failure, the licensee identified that the maintenance organizations improper procedure use and adherence was an extent of cause (condition adverse to quality). The licensee credited actions in another root cause evaluation to correct the identified extent of cause, however the actions Page 7 of 17
 
1Q/2013 Inspection Findings - River Bend 1 taken did not address the maintenance organizations procedure use and adherence issue. The licensee entered this issue into the corrective action program as Condition Report CR-RBS-2012-07250.
The failure to promptly identify and correct the maintenance organizations improper procedure use and adherence issue was a performance deficiency. The performance deficiency is more than minor, and is therefore a finding, because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, the inspector determined that the finding is of very low safety significance (Green). because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action component, because the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes.
Inspection Report# : 2012012 (pdf)
Significance:        Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control of the Control Building Chilled Water System The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the failure to maintain temperature control of the safety-related battery rooms. An engineering evaluation to change a procedure to allow gagging open of the control building heating and ventilation system control temperature valves failed to consider the appropriate environmental temperature limits for the rooms. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-07353.
The failure to maintain temperature control of the safety-related battery rooms was a performance deficiency. This performance deficiency is more-than-minor and is therefore a finding because it is associated with the design control attribute of the mitigating systems cornerstone and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, during a loss of offsite power with low seasonal temperatures, the gagged-open temperature control valve would reduce the battery rooms temperatures below their environmental design temperature and adversely affect the capacity of the safety-related batteries. In accordance with NRC Inspection Manual Chapter 0609, , Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, Section A.1, this finding screened as very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality. The engineering evaluation that changed the proper battery room controls was performed in 1997. Therefore, the finding did not have a cross-cutting aspect because the failed review is not indicative of current licensee performance.
Inspection Report# : 2012005 (pdf)
Significance:        Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Lubrication of the Standby Liquid Control Pump Motor Bearings Page 8 of 17
 
1Q/2013 Inspection Findings - River Bend 1 The inspectors identified a non-cited violation of Technical Specification 5.4.1.a for not establishing appropriate lubrication procedures for the standby liquid control pump motor bearings. Specifically, the station incorrectly used the Electrical Power Research Institute (EPRI) guidance for maintenance procedure by adding twice the amount of grease required. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-05573.
The failure to establish appropriate lubrication procedures is a performance deficiency. This performance deficiency is more-than-minor and is therefore a finding because if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, if the work instructions were not corrected, future work activities that grease the motor bearings in accordance with those work orders would over-grease the bearings, which may result in common-cause failures of standby motors. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, Section A.1, this finding screened as very low safety significance (Green). Specifically, the finding is a deficiency that affected the qualification of the standby liquid control pump motors; however, the systems maintained their operability. Because the most significant causal factor of the performance deficiency was station personnel and management failing to fully evaluate the previously identified inadequate lubrication of motors, this finding has a problem identification and resolution cross-cutting aspect associated with the corrective action program component [P.1(c)].
Inspection Report# : 2012005 (pdf)
Significance:        Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure for Lifting Leads Results in Inoperability of Standby Service Water Fan The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a due to a failure to follow work order instructions. Specifically, station personnel failed to follow the requirements of Procedure GMP-0042, Lifted Leads and Jumpers, Revision 13 when removing and reinstalling a time-delay relay for a standby service water cooling fan. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-06325.
The failure to follow work order instructions is a performance deficiency. This performance deficiency is more-than-minor because it is associated with the equipment reliability attribute of the mitigating systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure the correct wiring to the standby service water fan time-delay relay resulted in the inability of the fan to be started locally, which is required for remote shutdown of the plant. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, Section A, question 3, this finding required a detailed risk evaluation because the finding represented an actual loss of function of at least a single train for greater than the technical specification allowed outage time. The risk of the condition was evaluated by a senior reactor analyst. The sequence that would result in a risk increase is control room abandonment with concurrent maintenance being performed on the alternate bank of 5 fans. This would leave only 4 functional fans in one division of standby service water, whereas 5 fans are needed per design to meet the safety function.
The frequency of control room abandonment is approximately 5E-5/yr and the frequency of maintenance performed on one bank of standby service water fans is approximately 1E-2. Therefore, the frequency of a scenario where the failure of one fan to operate from the alternate shutdown panel would cause a measurable effect on risk is approximately 5E-7/yr. The other division of standby service water fans was unaffected by this condition.
Accordingly, the significance of the performance deficiency was determined to be very low (Green). This finding has a human performance cross-cutting aspect associated with the work practices component in that the electricians failed Page 9 of 17
 
1Q/2013 Inspection Findings - River Bend 1 to use adequate human error prevention techniques [H.4(a)].
Inspection Report# : 2012005 (pdf)
Significance:        Oct 11, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct Spurious Isolations of Reactor Core Isolation Cooling System The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality.
Specifically, the licensee failed to identify and correct an inadequate design of the reactor core isolation cooling (RCIC) system that resulted in spurious system isolations during main turbine trips. In response, the licensee installed a time delay into the circuit that had tripped the RCIC steam supply before the RCIC received a start signal. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-03439.
The performance deficiency was more than minor because it was associated with the design control 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, in that the repeated spurious isolations adversely affected the RCIC system reliability. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, this finding screened to a detailed risk evaluation which determined that the finding was of very low safety significance (Green). This finding does not have a cross-cutting aspect because the apparent cause of this finding was the licensees decision in 2008 to not add a time delay to the high differential pressure trip, and the NRC does not consider that cause to be representative of current licensee performance (Section 4OA5.2.a).
Inspection Report# : 2012010 (pdf)
Significance:        Oct 11, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Declare Reactor Core Isolation Cooling System Inoperable The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to declare the RCIC system inoperable when the system was unreliable for an automatic start following a main turbine trip. The licensee addressed the underlying safety concern by installing a time delay into the circuit that had tripped the RCIC steam supply before RCIC received a start signal. The licensee entered the finding into the corrective action program as Condition Report CR-RBS-2012-06015.
The performance deficiency was more than minor because it affected 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. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, this finding screened to a detailed risk evaluation which determined that the finding was of very low safety significance (Green). Because the most significant causal factor of the performance deficiency was that the organization had used the absence of information to determine RCIC operability, this finding has a cross-cutting aspect in the human performance area associated with the decision-making component, because the licensee had failed to demonstrate that the proposed action was safe in order to proceed rather than a requirement to demonstrate that it was unsafe in order to disapprove the action [H.1(b)] (Section 4OA5.2.b).
Inspection Report# : 2012010 (pdf)
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1Q/2013 Inspection Findings - River Bend 1 Significance:        Oct 11, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Prrevent Conflicts of Duty for Fire Brigade Members The inspectors reviewed a self-revealing, non-cited violation of License Condition 2.C.(10) because the licensee failed to prevent conflict of duties for fire brigade members, which affected the timely response to fires. In response, the control room initiated a night order to ensure that when a fire brigade member is called for fitness-for-duty testing, the staff will either designate a relief fire brigade member or arrange a deferral of the fitness-for-duty testing. The licensee plans to address long-term corrective actions through appropriate procedure changes at the fleet level. The licensee entered the finding into the corrective action program as Condition Report CR RBS-2012-03817.
The performance deficiency was more than minor because it was associated with the protection against external events 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. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, this finding screened to very low safety significance (Green) because the affected fire brigade member was unavailable for less than two hours. Because the most significant causal factor of the performance deficiency was that the licensee failed to ensure that conflicts between the fitness-for-duty and fire brigade procedures had been properly resolved prior to implementation, this finding has a human performance cross cutting aspect associated with resources because the licensee did not ensure that procedures were complete and accurate to assure nuclear safety [H.2(c)] (Section 4OA5.8).
Inspection Report# : 2012010 (pdf)
Significance:        Sep 28, 2012 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions to Ensure Reliability of the 480 VAC Molded Case Circuit Breakers and Unitized Motor Starters The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct within a reasonable period conditions adverse to quality associated with testing safety-related molded-case circuit breaker and unitized motor starter circuit breakers. The licensees immediate corrective actions included increasing the rate of breaker preventive maintenance and testing to reduce the long-standing risk-significant breaker backlog. The station documented the finding in Condition Report CR-RBS-2012-06364.
The performance deficiency was more-than-minor and is therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the safety-related molded-cased circuit breakers to respond to initiating events to prevent undesirable consequences. Specifically, failures of the affected breakers represent an increase in risk to safe plant operations, because to isolate a fault caused by a defective 480VAC breaker, the upstream feeder breaker would trip, thus causing a loss of power to additional safety-related components. Using Inspection Manual Chapter 0609, Appendix A, the finding is associated with the loss of mitigation equipment (Service Water pumps A and C), and so screened to a detailed risk evaluation. That evaluation determined that the incremental conditional core damage probability (ICCDP) was 2.1E-8 for a fire in one of the standby cooling tower electrical rooms, resulting in a loss of one train of service water pumps (A and C, or B and D), as a consequence of the failure of the proximate 480 VAC breaker to open. The risk was low because normal service water would be unaffected by the fire, and it would be unlikely that offsite power would be lost concurrently. The fire could also affect control room ventilation, but the analyst qualitatively concluded that this would not add more than negligibly to the overall risk. Consequently, the finding has very low safety significance (Green). The inspectors determined that the apparent cause of the finding was Page 11 of 17
 
1Q/2013 Inspection Findings - River Bend 1 a combination of two factors related to resources: station management did not ensure that each work group completed its actions to support timely resolution, and personnel vacancies from key positions hampered completion of the breaker testing program. The inspectors therefore determined the finding had a cross-cutting aspect in the human performance area associated with the resources component because station management did not ensure personnel resources were available to minimize long-standing equipment issues. [H.2(a)].
Inspection Report# : 2012004 (pdf)
Significance:        Sep 28, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Appropriately Tune the Reactor Core Isolation Cooling Turbine Speed Controller The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with inadequate instructions for tuning the reactor core isolation cooling (RCIC) terry turbine speed governor. The licensees immediate corrective actions included revising the maintenance procedure and recalibrating the RCIC turbine speed controller. The station documented the finding in Condition Reports CR-RBS-2012-01750 and CR-RBS-2012-01904.
This performance deficiency is more-than-minor and is therefore finding because it is associated with the equipment performance 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, during operation, this performance deficiency resulted in improper tuning of the turbine speed control system, which caused the turbine exhaust check valve to repeatedly slam against its open and shut valve stops and abnormally large turbine governor valve oscillations. Because the licensee had not tuned the turbine speed control system to run at a steady speed, the licensee removed RCIC from service to properly calibrate the control system, thereby adversely affecting RCIC availability. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification; did not represent a loss of system and/or function, did not represent either an actual loss of function of at least a single train for greater than its Technical Specification Allowed Outage Time, or two separate safety systems out-of-service for greater than its Technical Specification Allowed Outage Time; and did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined the apparent cause of this finding was the licensees failure to incorporate industry and vendor operating experience into the work instructions on February 12, 2011, to correct RCIC governor valve oscillations. Therefore, this finding has a cross-cutting aspect in the area of problem identification and resolution associated with the operating experience component because the licensee did not implement and institutionalize industry knowledge, including vendor recommendations, to support plant safety [P.2 (b)].
Inspection Report# : 2012004 (pdf)
Significance:        Jun 29, 2012 Identified By: NRC Item Type: NCV NonCited Violation High Pressure Core Spray Diesel Generator Bearing Lubrication Deficiencies The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failing to correct a condition adverse to quality for lubricating the high pressure core spray diesel generator bearings.
The station documented the finding in Condition Report CR-RBS-2012-02666.
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1Q/2013 Inspection Findings - River Bend 1 This performance deficiency was more than minor and was a finding because, if left uncorrected, inadequate lubrication work instruction could cause bearing failure due to inadequate lubrication or generator winding failure due to grease intrusion into the electrical windings in the generator. The significance of this finding was evaluated using a Phase 1 significance determination process screening and was determined to be of 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 seismic, flooding, or severe weather initiating events. The apparent reason the initial condition report was closed without correcting the work instruction to lubricate the high pressure core spray diesel generator bearings was that personnel who prepared and approved the operability evaluation were focused on proving operability not correcting a condition adverse to quality. Their focus was specific to the components ability to perform its function and not on completely identifying the issue in the corrective action program. Therefore, the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the station did not identify this issue completely, accurately, and in a timely manner commensurate with its safety significance [P.1(a)].
Inspection Report# : 2012003 (pdf)
Significance:      Jun 29, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Specify Manual Actions for Safety Relief Valve Operations During a Station Blackout Event The inspectors identified a non-cited violation of 10 CFR 50.63, Loss of All Alternating Current, paragraph (a) (2),
which states, in part, The reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. The capability for coping with a station blackout of specified duration shall be determined by an appropriate coping analysis. Licensees are expected to have the baseline assumptions, analyses, and related information used in their coping evaluations available for NRC review. Specifically, from November 1985 to May 17, 2012, the licensee failed to specify actions while ac power is unavailable to ensure that safety relief valves provided sufficient capacity and capability to ensure appropriate containment integrity is maintained during a station blackout event. This violation has been entered into the corrective action program as Condition Report CR-RBS-2012-03376.
The inspectors determined that failure to specify actions for safety relief valve operation in procedures in accordance with NUMARC-8700 was a performance deficiency. The finding was more than minor because it adversely affected the procedure quality attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to respond to undesirable consequences. Specifically, the station blackout coping procedures did not specify actions that would ensure the heat capacity temperature limit for the suppression pool would not be exceeded during the station blackout coping period. Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the inspectors determined that the Mitigating Systems Cornerstone was affected because the finding could cause degradation of core decay heat removal. Using Table 4a from the Phase 1 worksheet, the inspectors determined that the finding represents a loss of safety function; therefore, a Phase 2 analysis was necessary. However, the inspectors determined that a Phase 2 analysis was not sufficient to assess significance because of the complexity of the finding. Therefore, a Phase 3 analysis was necessary. The result of the Phase 3 analysis determined that the change in core-damage-frequency (?CDF) for the performance deficiency was 2.4E-7 or very low safety significance (Green). The senior reactor analyst determined that the change in large-early-release-frequency (?LERF) was 4.8E-8 or very low safety significance (Green). No cross-cutting aspect was identified because the most significant contributor was not indicative of current licensee performance (Section 4OA5).
Inspection Report# : 2012003 (pdf)
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1Q/2013 Inspection Findings - River Bend 1 Barrier Integrity Significance:        Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Effective Corrective Actions for Defects in MasterPact Breakers The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to promptly correct a condition adverse to quality. Specifically, station personnel failed to implement repairs to the mechanism-operated contact linkages for safety-related breakers, ultimately resulting in the failure of standby gas treatment filtration train 1B to start on demand. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-005894.
The failure to correct a condition adverse to quality is a performance deficiency. This performance deficiency is more-than-minor because it is associated with the systems, structures, and components and barrier performance attributes of the barrier integrity cornerstone and adversely affected the 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. Specifically, the standby gas treatment exhaust filter train failed to start during a surveillance test because of a nonconforming mechanical linkage in the feeder breaker resulting in unavailability for standby gas train 1B. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 3, Section C, question 1, the finding screened as very low safety significance (Green), because the finding represented only a degradation of the radiological barrier function provided by the standby gas treatment system. No cross-cutting aspect was assigned to this finding because the NRC concluded the finding did not reflect current licensee performance.
Inspection Report# : 2012005 (pdf)
Significance:        Sep 28, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Improper Hydrogen Igniter Breaker Trip Coil Setting The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the design bases for the power supply for the hydrogen igniter system into procedures used to set the associated power system supply breaker trip coil. The licensees immediate corrective actions included evaluating the proper trip coil setting and adjusting the trip coil accordingly. The station documented the finding in Condition Report CR-RBS-2012-02623.
This performance deficiency is more-than-minor and is therefore a finding because it is associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, this performance deficiency resulted in an incorrect trip coil setting, which decreased the reliability of the hydrogen igniters, which burn hydrogen in a controlled manner to prevent containment damage. Using Inspection Manual Chapter 0609.04, Initial Characterization of Findings, the finding required a significance evaluation per Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, because the unavailable Division 1 hydrogen igniters represented a degraded condition affecting containment barrier integrity that can potentially increase large early release frequency (LERF) without affecting the core damage frequency (CDF). Inspectors determined that this was a type B finding. Using section 6.0, the inspectors determined that the finding was of very low safety significance (Green) because the hydrogen igniters are arranged in two independent divisions such that each containment region has two igniters, one Page 14 of 17
 
1Q/2013 Inspection Findings - River Bend 1 from each division, controlled and powered redundantly so that ignition would occur in each region even if one division failed to energize. The inspectors determined that the apparent cause of this finding was that in response to earlier failures of the trip coil, the licensee had not investigated the problem thoroughly enough to identify and correct this performance deficiency. However, because the earlier failures had all occurred more than seven years ago, the inspectors determined that this cause did not reflect present licensee performance, so the inspectors did not assign a cross-cutting aspect to it.
Inspection Report# : 2012004 (pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Mar 30, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Request Briefings of the Dose Rates in High-Radiation Areas Before Entry The inspectors reviewed two examples of a self-revealing, non-cited violation of Technical Specification 5.7.1 that resulted because individuals failed to request briefings of the dose rates in high-radiation areas before entry. In response, the licensee coached the involved individuals involved about the acceptable radiation work practice. The licensee entered this issue into their corrective action program as Condition Reports 2012-07643 and 2013-01275.
The failure to request briefings of the dose rates in high-radiation areas before entry was a performance deficiency.
The significance of the performance deficiency was more-than-minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because the failure exposed workers to higher than anticipated radiation dose rates. The Occupational Radiation Safety Cornerstone was affected; therefore, the inspectors used Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, to determine the significance of the violation.
The violation had very low safety significance because: (1) it was not an as low as is reasonably achievable finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation had a cross-cutting aspect in the human performance area, associated with the work practices component, because licensee personnel failed to use human error prevention techniques, such as self- and peer-checking, commensurate with the risk of the assigned task such that work activities were performed safely [H.4(a)].
Inspection Report# : 2013002 (pdf)
Significance:        Mar 30, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of a Radiation Protection Technician to Provide Adequate Job Coverage The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.7.1 that resulted because a radiation protection technician failed to provide adequate job coverage. In response, the licensee coached the involved individuals involved about the acceptable radiation work practice. The licensee entered this issue into their corrective action program as Condition Report 2013-00479.
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1Q/2013 Inspection Findings - River Bend 1 The failure to provide adequate radiation protection job coverage was a performance deficiency. The requirement not met was Technical Specification 5.7.1. The significance of the performance deficiency was more-than-minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because the failure exposed workers to higher than anticipated radiation dose rates. The Occupational Radiation Safety Cornerstone was affected; therefore, the inspectors used Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, to determine the significance of the violation. The violation had very low safety significance because: (1) it was not an as low as is reasonably achievable finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation had a cross-cutting aspect in the human performance area, associated with the decision making component, because licensee personnel did not make a risk-significant decision using a systematic process when faced with uncertain or unexpected plant conditions [H.1 (a)].
Inspection Report# : 2013002 (pdf)
Significance:      Mar 30, 2013 Identified By: Self-Revealing Item Type: FIN Finding Failure to Provide Adequate Work Instructions for Installing Reactor Water Cleanup Pump Seals The inspectors reviewed a self-revealing finding associated with the licensees failure to provide adequate instructions for installing a new seal cartridge in the reactor water cleanup A pump. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-09015. In that condition report, the licensee developed a corrective action to revise all reactor water cleanup procedures and model work orders to verify proper installation of the pump seal.
The failure to provide adequate instructions for properly installing reactor water cleanup pump seal cartridges was a performance deficiency. The performance deficiency was more-than-minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and affected the cornerstone objective in that it caused increased collective radiation dose for occupational workers. Additionally, the finding was similar to example 6(i) in Appendix E to Manual Chapter 0612, "Power Reactor Inspection Reports -
Examples of Minor Issues." Using Manual Chapter 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," dated August 19, 2008, the inspectors determined the finding had very low safety significance because, although the finding involved ALARA planning and work controls, the licensee's latest three-year rolling average collective dose was less than 240 person-rem. This finding had a cross-cutting aspect in the human performance area, associated with the resources component, because the licensee failed to use complete, accurate and up-to-date procedures and work orders to perform the seal installation, which resulted in unnecessary dose [H.2(c)].
Inspection Report# : 2013002 (pdf)
Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission Page 16 of 17
 
1Q/2013 Inspection Findings - River Bend 1 has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
Miscellaneous Significance:      Oct 11, 2012 Identified By: NRC Item Type: NCV NonCited Violation Loss of Onsite Safety Review Committee Independence The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, for several examples of failures to follow Procedure EN OM 119, Onsite Safety Review Committee, Revision 8, which indicated that the onsite safety review committee failed to accomplish an independent review of station activities in accordance with the procedure. In response to this finding, the licensee developed a process to document the committee findings and reinforced roles and responsibilities for committee conduct, and committee members reviewed the implementing procedure. The licensee entered this finding into the corrective action program as Condition Report CR-RBS-2012-03739.
The multiple failures to follow the onsite safety review committee implementing procedure were performance deficiencies that were more-than-minor because failure to correct these performance deficiencies could compromise the nuclear safety oversight function of the committee, which could result in inappropriate decision-making on activities important to nuclear safety. In accordance with NRC Inspection Manual Chapter 0609 Appendix M, Significance Determination Process Using Qualitative Criteria, the finding was of very low safety significance because the performance deficiency did not result in any risk-significant issues. Because the most significant causal factor of the performance deficiency was the licensees failure to properly define, communicate and implement the roles for decision-making that affected nuclear safety, this finding has a human performance cross-cutting aspect associated with decision-making because the licensee failed to adequately communicate the authority and roles of the onsite safety review committee to the members [H.1(a)] (Section 4OA5.7).
Inspection Report# : 2012010 (pdf)
Last modified : June 04, 2013 Page 17 of 17
 
2Q/2013 Inspection Findings - River Bend 1 River Bend 1 2Q/2013 Plant Inspection Findings Initiating Events Significance:      Jun 29, 2013 Identified By: Self-Revealing Item Type: FIN Finding Failure to Establish Effective Preventive Maintenance for Components Used in High Critical Applications The inspectors reviewed a self-revealing finding for the failure to establish an adequate preventive maintenance strategy for the reactor feedwater regulating valves that resulted in several unplanned power changes due to packing steam leaks. In response, the station polished the pitted and scored valve stems and created a four-year periodic preventive maintenance task to replace the valve stems. The licensee entered this}}

Latest revision as of 13:53, 29 November 2024

2017 Q1-Q4 ROP Inspection Findings
ML20311A424
Person / Time
Site: River Bend Entergy icon.png
Issue date: 11/06/2017
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Office of Nuclear Reactor Regulation
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References
Download: ML20311A424 (717)
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