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05000285/FIN-2009006-06Failure to Update Intake Structure Design2009Q4The team identified a Severity Level IV, noncited violation for failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e).Specifically, the licensee failed to update Section 9.8, Raw Water Systems, of the Fort Calhoun Station Updated Safety Analysis Report after constructing a sheet pile alignment wall alongside the intake structure in 1982. Furthermore, this modification removed the slope from the river bottom. Additionally, recent sounding records indicate the river bottom near the intake structure is approximately the same depth as the center of the channel, thus, invalidating the updated safety analysis report statement. The licensee entered this condition into the corrective action program as CR 2009-3927.The finding is more than minor because the finding is determined to have a material impact on safety. Specifically, with the new sheet pile alignment wall, it could lead to a barge strike that is different than described in the updated safety analysis report. Using Supplement I of the NRC Enforcement Policy, this finding will be treated as a Severity Level IV violation. This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance
05000285/FIN-2009007-01Failure to Perform an Operability Evaluation of a Degraded Condition2009Q2The team identified a Green non-cited violation for the licensees failure to meet 10 CFR Part 50, Appendix B, Criterion V in that the licensee failed to perform an operability determination for a degraded condition. The licensee determined that certain relays classified as Functional Importance Determination 1, should be replaced every 9or 15 years depending on the duty cycle and environmental conditions. Most of the relays in the emergency diesel generator had been in service since initial installation, over 35 years ago. Subsequent to the inspection, the licensee performed an operability determination that showed all the effected relays were operable. This condition has been entered into the licensees corrective action program as Condition Reports 2009-2319 and 2342.The finding was determined to be greater than minor because the performance deficiency is associated with the procedure quality attribute (maintenance procedures)of the mitigating system cornerstone, and the performance deficiency adversely affected the associated cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, Phase 1 Significance Determination, and determined that it was of very low safety significance (Green) because the failure to perform the operability determination did not result in loss of operability or functionality and because the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of human performance, decision-making, in that the licensee did not make safety-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions to ensure safety is maintained H.1(a
05000285/FIN-2009007-02Failure to Perform Vendor and Industry Recommended Testing on Safety-Related and Risk Significant 4160 and 480 V Circuit Breakers2009Q2The team identified an unresolved item associated with inadequate maintenance procedures for 4160 and 480 V safety-related breakers. The team determined that maintenance procedures used to ensure that 4160 and 480 V safetyrelated breakers were being maintained and overhauled in a timely manner were inadequate. The licensee had no engineering analysis or technical basis to justify the deviation from vendor/Electric Power Research Institute guidance. At the end of the inspection, the licensee identified approximately 20 breakers that had failed over the last15 years and the team was waiting for additional information to determine if the failures were related to the inadequate maintenance. The team identified that the licensee was not performing the maintenance on the breakers as recommended by the vendor or Electric Power Research Institute guidelines. The licensee had completed a review of its breaker maintenance programs in November 2007 and modified it based on Electric Power Research Institute Documents TR-106857-V2 and TR-106857-V3, which are preventive maintenance program bases for low and medium voltage switchgear. The licensee only implemented portions of the recommended maintenance program, and had no engineering analysis or technical basis to justify the changes. Additionally, the guidance states in part that, this program assumes breakers are in nominally good condition to begin with. Breakers that have not been serviced for a very long time may need an overhaul or have a detailed inspection performed before this program is applied. The licensee had not been performing the entire vendor or Electric Power Research Institute recommended tests, inspections, and refurbishments on the breakers since installation. The team reviewed the licensee\'s circuit breaker maintenance procedures and records. The team determined that the licensee had not refurbished Asea Brown Boveri 4160 or General Electric 480 V safety-related and risk significant non-safety-related circuit breakers within the vendor specified 10-year maximum overhaul periodicity or the Electric Power Research Institute guidance of 12 years and had no engineering basis or evaluation to justify the deviation. The team compared the Electric Power Research Institute guidance and vendor-recommended maintenance requirements against the licensee\'s maintenance procedures and found that the licensee was not performing some of the recommended activities or had extended the periodicity of some inspections beyond even the Electric Power Research Institute recommended guidelines. The Fort Calhoun Station program for medium and low voltage switchgear and circuit breakers did not include most of the recommended testing and trending. Specifically, no testing of the operation of the 125-V DC control circuitry was performed at the voltages postulated to exist at the device terminals during design basis events. Contemporary industry standards and Electric Power Research Institute guidance recommend reduced control voltage testing as part of breaker maintenance. Vendor overhaul procedures include reduced control voltage testing on the as-found and as-left control circuit. While there is not an explicit requirement to perform reduced voltage testing on breaker control circuitry, the Electric Power Research Institute guidance recommends reduced voltage testing on breaker control circuitry in order to have reasonable assurance of reliable operation of control circuitry at the postulated minimum control voltage. Additional recommended testing per the preventative maintenance program basis DocumentsTR-106857-V2 and TR-106857-V3 that were not being performed included: Thermography inspections of the breakers and switchgear at recommended periodicity and trending, and: Measurement of the electrical resistance of coils and relays, trended over time to detect progressive failure of winding insulation and give an indication of the condition of these electrical devices. As a result, the team requested the basis for not performing all of the recommended maintenance activities. The licensee was unable to produce an engineering evaluation that allowed the use of the Electric Power Research Institute guidance versus the vendor guidance. Additionally, the team found that the licensee failed to update their in-use guidance when operating experience or new vendor information were issued. Because the licensee was unable to produce documentation demonstrating recommended maintenance had been performed at the appropriate intervals or which qualified the practice of extending the maintenance and refurbishment intervals, the team was concerned about the reliability of the safety-related and safety significant breakers that had not been overhauled within 10 years.n The licensee stated that the 10-year vendor requirement was based on breakers manufactured and lubricated with petroleum-based grease and that their Asea Brown Boveri circuit breakers were lubricated with synthetic-based grease, Anderol 757, which does not dry out as fast and extends the useful life of the lubrication. The licensee cited a May 11, 1995, letter from Asea Brown Boveri/Combustion Engineering that implied grease hardening was not an issue with Anderol 757 lubricant. The team identified operating experience which showed that other licensees had experienced grease hardening in Asea Brown Boveri breakers that contained the Anderol 757.Following the10 CFR Part 21 report issued by D. C. Cook on March 3, 1989, Asea Brown Boveri established the 10 year overhaul frequency. This report was issued after two Asea Brown Boveri 4160 V breakers failed to close because of hardened grease in their operating mechanism. Additional operating experience from Perry supported that grease hardening can occur in less than ten years, pertaining to the 4160 V C residual heat removal (RHR) pump breaker. It stated in part, Various anomalies were identified during the process of disassembling the breaker, and the lubricant within the operating mechanism appears to be hardened. Based on the breaker serial number it was determined that this breaker would have used the synthetic lubricate. This provided further evidence that synthetic grease can degrade in less than 10 years. Asea Brown Boveri breaker historical industry data showed that the lubrication in the operating mechanism tended to harden within 10 years and that this condition can cause sluggish breaker operation. The issue was entered into the licensees corrective action program- 14 V Enclosure and was being evaluated under Condition Report 2009-2306. This issue is unresolved pending review of the causes of the breaker failures as related to the improperly performed maintenance (Unresolved Item 05000285/2009007-02)
05000285/FIN-2009007-03Managing Gas Accumulation in Emergency Core Cooling System, Decay Heat Removal, and Containment Spray System2009Q2The team identified an unresolved item concerning the licensees program to identify and manage gas accumulation in emergency core cooling, decay heat removal, and containment spray systems. Specifically, on April 30, 2009, the licensee identified that a section of piping was inappropriately excluded from the scope of its Gas Management Program. Based on this, the licensee was reviewing the program to determine if additional piping was excluded that could cause voided piping, thereby resulting in the inoperability of a safety-related system. In response to NRC Generic Letter 2008-08, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, dated January 11, 2008 (ML072910759), the licensee developed a program to manage gas accumulation in the identified systems. By letter, dated October 14, 2008, Omaha Public Power District described the results of its analyses and concluded that gas accumulation in safety systems was unlikely to create conditions adverse to safety at the Fort Calhoun Station. However, on April 30, 2009, while performing ultrasonic examination of system piping under Work Order QC-ST-HPSI-0001, the licensee identified a gas void on the suction line to high pressure safety injection Pump SI-2B, downstream of Valve HCV-349. In its review, Omaha Public Power District found that it had inappropriately omitted this section of piping from the scope of the Gas Management Program. The team noted that the Updated Safety Analysis Report, section 6.2, page 11 of 35, revision 34, stated, in part, that this section of piping was not necessary to meet the core cooling requirements. However, opening Valve HCV-349 is in the Emergency Operating Procedures, and could introduce the void into the suction piping of high pressure safety injection Pump SI-2B.When discovered, the licensee conservatively declared that section of high pressure safety injection suction piping inoperable and entered Technical Specifications 2.3(2)(e), a 24-hour Limiting Condition for Operation. The licensee took actions to immediately vent and fill that section of piping and declared the system operable. The licensee initiated Condition Report 2009-2069 to determine the cause of the event and to evaluate whether other sections of piping were inappropriately excluded from the scope of its analyses that could render safety-related systems inoperable. At the conclusion of this inspection, the licensee had not completed its reviews. This issue is unresolved pending further NRC review of the licensees Gas Management Program Basis to determine if similar sections of piping were inappropriately excluded such that gas voids could render safety-related systems inoperable (Unresolved Item05000285/2009007-03)
05000285/FIN-2009007-04Failure to report a potential defect of breaker trip bars per 10 CFR Part 212009Q2The team identified an unresolved item concerning the extent of a deviation originally discovered in a failed safety-related breaker. An inadequate evaluation of the deviation was performed that could result in an event or condition not being properly reported under 10 CFR Part 21, 10 CFR Part 50.72, 10 CFR Part 50.73or 10 CFR Part 73.71.Description. On August 24, 2007, safety-related Breaker MCC-4B1-B01, Pressurizer Backup Heaters Bank 3 Group 8 failed its instantaneous trip setting on one phase. The failure analysis determined the failure to be curvature of the trip bar, likely due to a material defect. This failure was a deviation as defined by 10 CFR Part 21 (a departure from the technical requirements included in a procurement document) and the licensees governing procedure SO-R-1, Reportability Determinations. In order for this deviation to be reportable under 10 CFR Part 21, 10 CFR 50.72, 10 CFR 50.73 or 10 CFR 73.71, the deviation must be determined to be a defect. As defined by 10 CFR Part 21, a defect includes deviations in a basic component delivered to a purchaser for use in a facility or an activity subject to the regulations in this part if, on the basis of an evaluation, the deviation could create a substantial safety hazard. In evaluating the deviation, the licensee arbitrarily determined that the deviation only applied to breakers with the same date code as the failed breaker. This conclusion was reached with no engineering basis and without consultation with the vendor of the breaker. In evaluating deviations, only the vendor can fully determine the extent of the deviation and its potential effect on other plant components. Since Procedure SO-R-1does not direct vendor notification unless the initial deviation is potentially associated with a substantial safety hazard, it was not possible to determine whether the deviation existed in other components. The licensee determined there were no other breakers with the same date code located anywhere on site, thus the only breaker assumed to have the deviation was the initial breaker that failed. Due to safety-related function of the particular breaker, it was determined that there was no substantial safety hazard, and the event was not reportable under 10 CFR Part 50.72 or 10 CFR Part 50.73. Thus the licensee determined that any reporting requirements required under Part 21 were satisfied, as described in 10 CFR Part 21.2(c). However, since the extent of the deviation was measured against breakers only with the same date code, and without consultation with the vendor, the evaluation was inadequate to determine if the event was reportable under 10 CFR Part 50.72 or 10 CFR Part 50.73. In addition, a proper evaluation of components stored in the warehouse could not be made resulting in an inadequate evaluation to determine if the condition was reportable under 10 CFR Part 21.On November 14, 2007, safety-related breaker MCC-3C1-B01, Pressurizer Backup Heaters Bank 2 Group 4 failed its 300 percent thermal test, instantaneous trip setting, on all three phases. This breaker was the same make and model as the breaker that failed on August 24, 2007, but was a different date code. The failure analysis of this breaker was documented in the same report as the initial breaker failure. While the failure mechanism of this breaker was different than the previous breaker failure, the failure analysis noted that the trip bar was curved, though it did not contribute to the failure. The first breaker failure was determined to be curvature of the trip bar, and the second breaker was exhibiting the same characteristics. Since the two failures occurred so close together in time and the failure analyses were documented in the same report, the licensee could have reasonably questioned the extent to which the deviation present in the first breaker occurred. After a review of the two breaker failure events, the team asked the licensee to determine if other breakers were installed in the plant or stored in the warehouse that contained the same deviation. This issue is unresolved pending review of potentially affected breakers after the licensee consults with the vendor to determine if a substantial safety hazard exits (Unresolved Item 05000285/2009007-04)
05000285/FIN-2010002-01Inadequate Reportability Guidance2010Q1The inspectors identified a Severity Level IV noncited violation of Fort Calhoun Technical Specification 5.8.1 for inadequate corrective action documents. Specifically, the documents do not adequately address assigning reportability evaluations. As a result, the licensee failed to evaluate the reportability of a condition that was determined to be reportable until questioned by the inspectors. The inspectors determined that the licensees inadequate corrective action documents were a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Policy. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was potentially affected. Specifically, the NRC relies on the licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore a finding. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated in accordance with the NRC Enforcement Policy. The finding was reviewed by NRC management and due in part to its repetitive nature the violation was determined to be of more than minor significance, however since it was not found to be willful, and was entered into the corrective action program, this violation is being treated as a Severity Level IV noncited violation consistent with the NRC Enforcement Policy
05000285/FIN-2010003-01Failure to Provide Adequate Limiting Condition for Operation for High River Level2010Q2The inspectors identified a noncited violation of 10 CFR 50.36(c)(2)(ii)(B) for the failure to include an adequate limiting condition for operation in the technical specification. Specifically, the reactor cannot be placed in a cold shutdown condition using normal operating procedures when the river level exceeds 1009 feet mean sea level, as required by Technical Specification 2.16. This violation has been entered into the licensee‟s corrective action program to determine the appropriate limiting condition for operation. The inspectors determined that the licensee‟s failure to include an adequate limiting condition for operation in the technical specification was a performance deficiency. This finding is more than minor because it affected 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 to prevent undesirable consequences, in that the reactor cannot be placed in a cold shutdown condition using normal operating procedures when the river level exceeds 1009 feet mean sea level. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609, Appendix A, to determine its significance. Using Attachment 1 of that appendix, the inspectors determined that this finding had very low risk significance because the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event, using the criteria included in Exhibit 1 of Inspection Manual Chapter 0609 Attachment 4. Since the finding is not indicative of current licensee performance, there is no crosscutting area assigned to this finding.
05000285/FIN-2010003-02Failure to conduct an adequate audit of emergency preparedness interfaces with offiste authorities2010Q2The inspectors identified a noncited violation of 10 CFR 50.54(t)(2) for the failure to conduct an evaluation of the adequacy of interfaces between the licensee and state and local governments during a periodic review of the site emergency preparedness program. Specifically, the quality assurance audit team, for the February 2010 emergency preparedness audit, did not evaluate the adequacy of interfaces with offsite agencies and did not contact state or local emergency management or radiological health agencies during the audit to obtain information about their working relationships with the licensee. The licensee has placed this violation in their corrective action program as Condition Report 2010-2078. This finding is more than minor because it affected the offsite emergency preparedness attribute of the Emergency Preparedness Cornerstone objective. This finding was determined to be of very low safety significance because it was a failure to comply with an NRC requirement and was not associated with the planning standards of 10 CFR 50.47(b). This finding is associated with the resources component of the human performance crosscutting area (H.2(b))
05000285/FIN-2010003-03Failure to Conduct Drills to Maintain Environmental Monitoring Skills2010Q2The inspectors identified a noncited violation of 10 CFR 50.47(b)(14) for the failure to conduct drills that were adequate to maintain key skills. Specifically, environmental monitoring teams were not required to collect environmental samples during the 2008 and 2009 annual environmental monitoring drills. The licensee has placed this violation in their corrective action program as Condition Report 2010-2055. This finding is more than minor because it affected the emergency response organization performance and procedure quality cornerstone attributes of the Emergency Preparedness Cornerstone objective. The finding is of very low safety significance because it is a failure to comply with NRC requirements, was associated with nonrisk significant planning standard 10 CFR 50.47(b)(14), and was not a functional failure of the planning standard. This finding is associated with the resources component of the human performance crosscutting area (H.2(c))
05000285/FIN-2010003-04Protective Action Recommendation processes allow for the unnecessary evacuation of the Public2010Q2The inspectors identified a noncited violation of 10 CFR 50.47(b)(10) and 50.54(q) for the failure to develop and put into place guidelines for the choice of protective actions during an emergency that were consistent with federal guidance. Specifically, the licensee‟s methodology for determining protective action recommendations could result in recommendations to evacuate members of the public in areas where dose projections did not exceed EPA protective action guides. The licensee has placed this violation in their corrective action program as Condition Report 2010-2174. This finding is more than minor because it adversely affected the emergency response organization performance and procedure quality cornerstone attributes of the Emergency Preparedness Cornerstone objective. This finding was determined to be of very low safety significance because it was a failure to comply with NRC requirements, is a finding associated with a risk significant planning standard, and is not a risk significant planning standard functional failure or degraded function. This finding was associated with the operating experience component of the problem identification and resolution crosscutting area (P.2(a)) (Section 4OA1).
05000285/FIN-2010003-05Failure to Submit a Required Licensee Event Report2010Q2The inspectors identified a noncited violation for the failure to submit a licensee event report within 60 days of discovery of an event as required by 10 CFR 50.73. Specifically, the turbine-driven auxiliary feedwater pump, FW-10, was inoperable from February 26 until April 6, 2009, which is a reportable condition required by 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by technical specifications. On March 11, 2009, the electric motor-driven auxiliary feedwater pump, FW-6, was inoperable for approximately four hours when diesel generator 1 was inoperable. With both auxiliary feedwater pumps simultaneously inoperable, this was a reportable condition required by 10 CFR 50.73(a)(2)(v) as an event that could have prevented fulfillment of a safety function. The licensee entered this violation into their corrective action program, completed a reportability evaluation and determined that a licensee event report was required to be submitted within 60 days of April 6, 2009, and had not been submitted. The licensee event report will be submitted prior to August 10, 2010. The inspectors determined that the licensee‟s failure to submit a Licensee Event Report was a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was potentially affected. Specifically, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore a finding. The inspectors determined that this finding was not suitable for evaluation using the significance determination process and, as such, was evaluated for traditional enforcement only in accordance with the NRC Enforcement Policy. This is a Severity Level IV noncited violation consistent with Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy.
05000285/FIN-2010003-06Failure to Perform a Proper 50.59 Evaluation2010Q2On April 9, 2010, the licensee repaired a section of power cable for motor control center MCC-3A1 with cable splices. Approximately 17 feet of 500 MCM cable was removed from each of the three phases for the supply to MCC-3A1 and Burndy compression type butt splices were used to splice new cables to the remaining existing cables. The inspectors reviewed Section 8.5, Initial Cable Installation Design Criteria of the USAR. USAR 8.5 states, in part: The Cable and Conduit Schedule Notes, Figure 8.5-1, provides the standard design criteria for cables and conduits. Deviation from the standard criteria is acceptable provided an analysis has been completed which justified the deviation. USAR Figure 8.5-1, Cable and Conduit Schedule Notes, Note 19 states: Splicing in cable trays is not allowed unless specifically called for on drawings. Exceptions to this requirement shall require the written approval of the engineer. USAR Figure 8.5-1, Note 26 states: Deviations from the standards stated above is (are) acceptable provided an analysis has been performed to justify the deviation. USAR Section 8.5.4.c states: Cable splicing in cable trays is used only for connection of incoming and outgoing cables with containment electrical penetration conductors. The licensee performed a 50.59 Screen in accordance with the guidance provided in FCSG-23, 10 CFR 50.59 Resource Manual. The guidance adopts NEI 96-07, Revision 1 Guidelines for 10 CFR 50.59 Implementation which includes five screening questions to determine if a complete evaluation of 10 CFR 50.59 is required. The licensee determined that a cable splice was an equivalent replacement for cable, and thus it screened out in accordance with NEI-96-07 and no evaluation of 10 CFR 50.59 was required. The inspectors determined that a cable splice is not an equivalent replacement, thus a violation of 10 CFR 50.59 occurred for failure to perform an evaluation of the cable splice against the criteria set forth in 10 CFR 50.59. The violation would be greater than minor only if prior NRC approval was required. The inspectors are reviewing the technical aspects of this issue to determine if prior NRC approval would have been required. In accordance with the guidance in Inspection Manual Chapter 0612, an unresolved item is warranted if more information is required to determine if the performance deficiency is more than minor, URI 05000285/2010003-06, Failure to Perform a Proper 50.59 Evaluation
05000285/FIN-2010003-07Licensee-Identified Violation2010Q2Licensee Event Report 05000285/2009-004 identified that containment closure was violated and it was identified that the condition was reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) on December 16, 2009. The events described in licensee event report 05000285/2009-004 were reported within 60 days of December 16, 2009; however, the containment integrity violation which occurred on November 1, 2009, was not, despite the obvious similarity between the three events. The November 1, 2009, event was not reported until June 4, 2010; 170 days after the event was discovered. This is a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(i)(B) for failure to submit a required licensee event report within 60 days of a condition prohibited by technical specifications.
05000285/FIN-2010004-01Inadequate Documentation of the Adequacy of Design for the Pumps that Transfer Fuel Oil from Storage Tank FO-10 to FO-12010Q3The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III due to the failure of the licensee to perform suitable testing to determine the adequacy of the design of equipment related to transferring diesel fuel from one storage tank to another. Specifically, the inspectors questioned whether fuel oil transfer pump FO-37 or a portable hand pump to be used in the event that FO-37 was unavailable to transfer fuel from storage tank FO-10 to FO-1 would be able to perform the design function. No calculations or previous testing documentation could be provided and when tested to demonstrate that the portable hand pump could perform the intended design function, the portable hand pump failed. Subsequently, the licensee evaluated that fuel oil transfer pump FO-37 is adequately designed to transfer fuel oil from FO-10 to FO-1. The licensee entered this issue into the corrective action program as Condition Reports 2010-3123, 2010-3921, and 2010-4315. The inspectors determined that the licensees failure to provide calculations or testing documentation that fuel oil transfer pump FO-37 or the designated portable hand pump could perform the intended design function was a performance deficiency. This finding is greater than minor because it affected the Mitigating System Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the equipment performance attribute to maintain availability and reliability of the diesel generators. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has a very low safety significance (Green) because it was not a design or qualification deficiency, does not represent an actual loss of safety function nor did it screen as potentially risk significant for external events. Since the finding is not indicative of current licensee performance, there is no crosscutting aspect assigned to this finding (Section 1R15).
05000285/FIN-2010004-02Failure to Submit a Required Licensee Event Report2010Q3The inspectors identified a Severity Level IV noncited violation for the failure to submit a licensee event report within 60 days as required by 10 CFR 50.73. Specifically, the diesel fuel oil storage system was inoperable for approximately 24 hours from January 6, 2010, until January 7, 2010. On January 6, 2010, fuel oil transfer pump FO-37 was inoperable due to a fire main rupture submerging the pump for approximately 24 hours. With no other means to transfer fuel from storage tank FO-10 to FO-1, the fuel oil storage system was inoperable, and the fuel volume in FO-10 was unavailable. This was reportable condition required by 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by technical specifications. The licensee performed a reportability evaluation, and the violation was entered into the corrective action program as Condition Report 2010-3865. The inspectors determined that the licensees failure to submit a licensee event report was a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was potentially affected. Specifically, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore more than minor. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for traditional enforcement only in accordance with the NRC Enforcement Policy. This is a Severity Level IV violation as defined in Section 2.2.1.c of the NRC Enforcement Policy (Section 1R15).
05000285/FIN-2010004-03Failure to Update the Updated Safety Analysis Report Solid Waste2010Q3The inspectors identified a Severity Level IV, noncited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, paragraph (e) which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Contrary to the above, the licensee failed to update periodically the Updated Safety Analysis Report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Specifically, since December 2006, the licensee stored a significant source of radioactivity in the original steam generator storage facility but failed to describe the source, volume, and storage of radioactive equipment in the Updated Safety Analysis Report. The licensee has entered this violation into their corrective action program as Condition Report 2010-3636. The inspectors determined that the failure to update the Updated Safety Analysis Report as required by 10 CFR 50.71(e), Maintenance of Records, Making of Reports was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding was more than minor because it had a material impact on licensed activities in that a radioactive solid waste storage facility was relocated from the plant radiological controlled area to the owner controlled area without being described in the Updated Safety Analysis Report. The finding was characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy (Section 1R17).
05000285/FIN-2010004-04Failure to Translate Calculation into Calibration Procedure2010Q3The inspectors identified a 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, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, the licensee failed to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, since January 2009, the licensee failed to correctly translate results of Calculation FC 05561, CCW Relief Valve Setpoints, into calibration procedures used to calibrate pressure control switches PCS-412 and PCS-413. The licensee has entered this violation into their corrective action program as Condition Report 2010-3658. The inspectors determined that the failure to correctly translate the results of the setpoint calculation into calibration procedures and instructions as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control is a performance deficiency. The finding was more than minor because it adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Additionally, the finding was more than minor because the finding resulted in a condition where there was a reasonable doubt on the operability of the component cooling water system containment isolation valves. Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the finding was determined to have very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building. This finding has a crosscutting aspect in the area of human performance work practice because the licensee failed to define and effectively communicate expectations regarding procedural compliance and personnel following procedures. Specifically, in January 2009, the licensee failed to effectively communicate expectations regarding personnel following procedures to implement calculation changes (H.4(b))(Section 1R17).
05000285/FIN-2010004-05Failure to Perform a 10 CFR 50.59 Evaluation2010Q3The inspectors identified a Severity Level IV violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the facility as described in the Updated Safety Analysis Report. On April 9, 2010, the licensee changed the facility as described in the Updated Safety Analysis Report to install a cable splice in a safety related cable without determining if prior NRC approval was required. The licensee took actions to make the modification temporary until a permanent repair could be made and entered the issue into the corrective action program as Condition Report 2010-4466. Fort Calhoun Station utilizes NEI 96-07 as their process to meet 10 CFR 50.59 requirements. Their failure to perform a 10 CFR 50.59 evaluation, in accordance with NEI 96 07, prior to changing the facility as described in the Updated Safety Analysis Report is a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was potentially affected. Specifically, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore more than minor. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for Traditional Enforcement only in accordance with the NRC Enforcement Policy. The inspectors concluded that the 10 CFR 50.59 evaluation would have likely identified that prior NRC approval would have been required, unless the change to the facility was for a short duration of time. This was due to the introduction of additional potential failure mechanisms of the splices that are age-dependent. Since the licensee subsequently classified the cable splice as a temporary modification, and scheduled to be removed during the next refueling outage, the aging mechanisms would no longer be applicable. Therefore, this is a Severity Level IV violation as defined in Section 2.2.1.c of the NRC Enforcement Policy (Section 1R20).
05000285/FIN-2010004-06Failure to Follow Radiation Work Permit Requirements2010Q3The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.8.1, for failure to follow radiation work permit requirements. On November 13, 2009, two individuals became contaminated while cleaning the gasket seating surface on the endbell of the letdown heat exchanger because they did not use face shields as required by the radiation work permit. The licensee immediately restricted the two individuals from entry into the radiologically controlled area, conducted a coaching session with the individuals involved and placed this issue into the corrective action program as Condition Report 2009-5688. The failure to follow the instructions listed on a radiation work permit was a performance deficiency. 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 follow radiation work permit instructions increased personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work practices, human error prevention techniques, because the individuals failed to use self and peer checking to ensure they were signed onto the appropriate task for the work to be performed (H.4(a))(Section 2RS01).
05000285/FIN-2010004-07Failure to Properly Plan a Maintenance Activity2010Q3The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.8.1, for failure to appropriately control radiation exposures due to improperly planned maintenance activities associated with Work Package 09-AP-20. The maintenance work involved valve modifications and boric acid system cleanups. These activities resulted in exceeding the original dose estimate by more than 50 percent. The licensee entered this issue into the corrective action program as Condition Reports 2009-6171, 2009-6264 and 2010-1696. The failure to properly plan maintenance activities to minimize personnel radiation dose is a performance deficiency. This finding is greater than minor because it affected the Occupational Radiation Safety Cornerstone attribute of program and process in that ALARA planning or radiological controls did not prevent unplanned, unintended dose for a work activity. This caused increased collective radiation dose for the job activity to exceed the planned dose of approximately 14 rem by more than 50 percent. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this finding to be of very low safety significance because the finding involved ALARA planning and controls and the licensees latest rolling 3-year average does not exceed 135 person-rem. This finding had an associated human performance crosscutting aspect in the work practices component because the licensee did not ensure supervisory and management oversight of work activities, including the contractor, to maintain doses ALARA (H.4(c))(Section 2RS02).
05000285/FIN-2010004-08Inadequate Maintenance Procedure Results in Water in East Switchgear Room and Room 192010Q3The inspectors reviewed a self-revealing Green noncited violation of Fort Calhoun Station Technical Specification 5.8.1, for the licensees failure to provide an adequate maintenance procedure for fire protection system flushing. Specifically, while performing OP-PM-FP-1000 on August 19, 2010, water backed up the VA-87 drain line and spilled onto the east switchgear room floor, into Room 19 below, as well as pooling on top of and inside of cable trays. The licensee has entered this issue into their corrective action program as Condition Report 2010-4423. The inadequate maintenance procedure is a performance deficiency. This finding is more than minor because if left uncorrected the performance deficiency could have the potential to lead to a more significant safety concern. Specifically the use of OP-PM-FP-1000 allows the potential wetting of safety related equipment in the east switchgear room and Room 19. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609, Appendix A, to determine its significance. Using Attachment 4 of that appendix, the inspectors determined that the finding has 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. Through conversations with the fire protection system engineer and other licensee members and the fact that similar issues have occurred in the past, the inspectors determined that the primary cause of this finding was the failure to adequately assess the significance of previous condition reports which would have required them to perform a more thorough cause evaluation. Therefore, this finding has a crosscutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems, such that, the resolutions address causes and extent of conditions, as necessary (P.1(c))(Section 4OA2).
05000285/FIN-2010004-09Failure To Perform Vendor And Industry Recommended Testing On Safety-Related And Risk Significant 4160 V And 480 V Circuit Breakers2010Q3The inspectors identified a Green noncited violation of Technical Specification 5.8.1(a) for inadequate procedures associated with 4160 V and 480 V safety-related breaker maintenance procedures. The inspectors determined that maintenance procedures used to ensure that 4160 V and 480 V safety-related breakers were being maintained and overhauled in a timely manner were inadequate. The licensee did not have an engineering analysis or technical basis to justify the deviation from vendor and/or Electric Power Research Institute guidance. The inspectors determined that this issue affected the procedure quality attribute for maintenance procedures of the Mitigating System Cornerstone of reactor safety. Specifically, the issue was more than minor because the failure to incorporate the vendor required maintenance and frequency or fully incorporate Electric Power Research Institute maintenance recommendations for extending the service interval into maintenance procedures for safety related breakers. If left uncorrected, this failure affected the availability, reliability, and capability of mitigating systems that respond to initiating events to prevent undesirable consequences because the reliability of safety-related breakers refurbished using the deficient procedures cannot be predicted. This issue was entered into the licensees corrective action program as Condition Report 2009-2306. Using the Significance Determination Process, Phase 1 Screening Worksheet, for the Initiating Events, Mitigating Systems, and Barriers Cornerstones the finding was potentially risk significant for multiple systems. Because the probability of multiple system effects is not effectively addressed by a Phase 2 analysis, a Phase 3 analysis was performed. The analyst determined that while the licensee failed to perform adequate maintenance on the breakers, the actual failure rate of the breakers was no greater than the theoretical design failure rate. The finding was determined to be of very low safety significance because the deficiency did not result in any loss of function. The finding was not risk significant due to a seismic, flooding, or severe weather-initiating event and because other plant-specific analyses that identify core damage scenarios of concern were not impacted. This finding has a crosscutting aspect in the area of problem identification and resolution because the licensee did not effectively incorporate pertinent industry operating experience into the preventive maintenance programs for the 4160 V and 480 V safety-related and risk significant non-safety-related circuit breakers (P.2(b))(Section 4OA2)
05000285/FIN-2010004-10Inadequate Maintenance Procedure Results in Plant Shutdown2010Q3A self-revealing Green noncited violation of Fort Calhoun Station Technical Specification 5.8.1 occurred for an inadequate procedure for verifying the connection between cable lugs and cables. This inadequacy resulted in the loss of Motor Control Center MCC-3A1 and a subsequent plant shutdown. The licensee repaired the affected equipment and entered this issue into the corrective action program as Condition Report 2010-4423. The inspectors determined that the licensees inadequate maintenance procedure was a performance deficiency. This finding was greater than minor because it was similar to a non-minor example 4.b in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that a procedural error caused a reactor trip or other transient. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has very low safety significance because all of the items in Table 4a, of the Mitigating Systems Cornerstone checklist, were answered in the negative. Since the finding is not indicative of current licensee performance, there is no crosscutting aspect assigned to this finding (Section 4OA3).
05000285/FIN-2010004-11Licensee-Identified Violation2010Q3Technical Specification 5.8.1 requires written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, February 1978. Section 7.e of Appendix A to Regulatory Guide 1.33 requires radiation protection procedures. Procedure RP-307, Use and Control of Temporary Shielding, Revision 18, Step 5.3 states that, No temporary shielding shall be installed, removed or modified unless authorized. Step 7.4.4.c of this procedure states that, Radiation Protection personnel are NOTIFIED prior to removing shielding. Contrary to these requirements, on December 8, 2009, the containment coordinator removed ten lead shielding blankets hanging on a hand rail without notifying radiation protection. The removal of the blankets increased the dose rate on that side of the railing resulting in increased dose rates. The containment coordinator was counseled by the radiation protection supervisor and the ALARA coordinator. The inspectors determined this finding to be of very low safety significance because: (1) it did not involve ALARA - 55 Enclosure planning and controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This issue was entered into the licensee\'s corrective action program as Condition Report 2009 6454.
05000285/FIN-2010004-12Licensee-Identified Violation2010Q3Licensee Event Report 05000285/2010-004 identified that accelerometer flow elements for both pressurizer safety valves were inoperable from April 28 to June 2, 2010. This condition is prohibited by technical specifications after 7 days, therefore meeting the criteria for a condition prohibited by technical specifications on May 5, 2010. The licensee event report was submitted 17 days later, on August 30, 2010. This is a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(i)(B) for failure to submit a required licensee event report within 60 days of a condition prohibited by technical specifications.
05000285/FIN-2010005-01Failure to Maintain Licensed Operator Examination Integrity2010Q4The inspectors identified a Green noncited violation of 10 CFR Part 55.49, Integrity of Examinations and Tests, for the failure of the licensee to ensure that the integrity of an operating test administered to licensed operators was maintained. Two licensed operators received five job performance measures for their retake operating tests that had been potentially compromised during earlier weeks when this weeks operating test book was left out and uncontrolled overnight in the training building. These job performance measures were removed from the operating tests for subsequent weeks and a condition report was written to ensure that these job performance measures were not used in subsequent weeks. However, these actions did not prevent these job performance measures from being used for the retake operating tests for two licensed operators that failed previous operating tests. This resulted in a compromise of operating test integrity because control of these items was lost; however, it did not lead to an actual effect on the equitable and consistent administration of the examination. This issue was entered into the licensees corrective action program as Condition Report 2010-5977. The failure of the licensees training staff to maintain the integrity of examinations administered to licensed operations personnel was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it adversely impacted the human performance attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, it could have become more significant in that allowing untested licensed operators (in this case, operators that had the potential to have an invalid test because of the lack of examination integrity) at the controls could be a precursor to a more significant event if undetected performance deficiencies develop. Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process, the finding was determined to have very low safety significance (Green) because, although the finding resulted in a compromise of the integrity of operating test job performance measures and compensatory actions were not immediately taken when the compromise should have been discovered in 2009, the equitable and consistent administration of the exam was not actually impacted by this compromise. This finding has a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because the licensee did not take appropriate corrective actions to address safety issues in that an operating test compromise issue occurred that was entered into the corrective action program as Condition Report 2009-4066. This corrective action document stated that these compromised items shall not be used on any subsequent operating tests for that cycle and they were subsequently used on the 2009 annual operating test.
05000285/FIN-2010005-02Failure to Perform a Risk Assessment When Required by 10 CFR 50.65(a)(4) for Maintenance in the Vicinity of Safety-Related Equipment2010Q4The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for failing to perform a risk assessment prior to performing activities involving a man basket in the vicinity of the T1 transformer. The licensee has entered this performance deficiency into the corrective action program as Condition Report 2010-4689. The inspectors determined that the licensees failure to perform a risk assessment and implement appropriate risk management actions was a performance deficiency. The finding was more than minor because it was associated with the protection against external factors attribute of the Initiating Events Cornerstone. It directly 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. Additionally, if left uncorrected, the practice of not adequately evaluating crane activities in the vicinity of safety-related equipment by appropriately trained individuals would become a more significant safety concern. Specifically, in that it could result in a more than minimal increase in risk associated with other risk important equipment that would not be identified and not result in appropriate actions being taken. The inspectors evaluated this finding using the Appendix K, Maintenance Risk Assessment, and Risk Management Significance Determination Process worksheets of Manual Chapter 0609 because the finding is a maintenance risk assessment issue. Flowchart 1, Assessment of Risk Deficit, requires the inspectors to determine the risk deficit associated with this issue. This finding was determined to be of very low safety significance because the incremental core damage probability deficit was less than 1 x 10-6. Because of the confusion with performing a risk assessment with a crane but not with a man basket, the finding had crosscutting aspects in the area of human performance associated with resources in that the licensee failed to provide complete, accurate, and up-to-date procedures.
05000285/FIN-2010005-03Failure to Properly Apply an Approved ASME Code Case2010Q4The inspectors identified a Green noncited violation of 10 CFR 50.55a(b)(5)(i) because the licensee failed to adequately apply ASME Section XI Code Case N-513-2 when they evaluated a degraded section of raw water piping for operability. The licensee has entered this performance deficiency in the corrective action program as Condition Report 2010-5680. The inspectors determined that the licensees failure to adequately apply ASME Code Case N-513-2 was a performance deficiency. The finding was more than minor because it was associated with the protection against external factors 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 prevent undesirable consequences. Additionally, if left uncorrected, improper application of an approved code case would become a more significant safety concern in that it could result in the failure to identify inoperable safety related piping. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has a very low safety significance (Green) because it was not a design or qualification deficiency, does not represent an actual loss of safety function, nor did it screen as potentially risk significant for external events. Because the licensee revised an old operability determination and did not recognize that the code case application was incorrect, the finding had crosscutting aspects in the area of human performance associated with decision-making in that the licensee failed to make safetysignificant or risk-significant decisions using a systematic process.
05000285/FIN-2010005-04Calibration Failures of RPS Trip Units 6 and 72010Q4The inspectors identified an unresolved item concerning calibration of steam generator low-pressure trip unit A/TU-6 and asymmetrical steam generator transient (ASGT) trip unit A/TU-7. Further investigation is required to determine whether multiple performance deficiencies exist and if they are more than minor. On November 8, 2010, the licensee performed Surveillance Test IC-ST-RPS-0044, calibration of steam generator low-pressure trip unit A/TU-6 and asymmetrical steam generator transient (ASGT) trip unit A/TU-7. During the performance of this test out of tolerance as found values were recorded at approximately 10:30 a.m. Specifically, on Attachment 9.1 - Data Sheet 1, terminal 73 and terminal 75 were elevated out of tolerance by 0.0001 Vdc and 0.0002 Vdc respectively. These values normally correspond to RC-2A and RC-2B steam generator pressure. RC-2A and RC-2B pressure transmitters scale from 1-4 volts corresponding to 0 - 1000 psi. The condition observed was nonconservative, in that, increased voltage would mask a low pressure condition to the reactor protection system (RPS) by a linear amount. The intent of gathering this data is to ensure that there is minimal degradation of the signals by the circuitry prior to trip unit input. The remaining as found data required by the surveillance test was recorded by the instrumentation and controls (I&C) technician. The remaining values including trip unit A/TU-6 and trip unit A/TU-7 input values were in specification. The out-of-tolerance values failed the surveillance test. Work Request 157517 was generated to troubleshoot and repair. Condition Report 2010-5645 documented the out of tolerance values. During troubleshooting efforts it was determined the issue resided with the circuit. Specifically, terminal 74 is connected to common and should have a value of zero Vdc. Instead, this terminal was reading greater than zero Vdc by a few millivolts. Common to all three terminals is AI-31A-AW12 B2 contact module, which is part of the ASGT test circuit and should not affect the circuits. Instrumentation and controls technicians knew this module had previously been an issue. Condition Reports 200302822 and 2009- 2317 document past out of tolerance results. Cycling the contact module or replacing it had cleared out of tolerances values in the past, therefore part of the FC-1212 troubleshooting plan was to cycle the contact module. The FC-1212 was executed and no maintenance activities were performed. Surveillance Test IC-ST-RPS-0044 was performed again to check the required values for change. The out of tolerance values were now in tolerance. The on-shift I&C technician did not intend to complete the surveillance test. Instead, the trip units were left in bypass and the results were discussed with the shift manager including a safety concern regarding the contact module. This concern was documented in Condition Report 2010-5667 on November 8, 2010, at 3:00 p.m. The condition report questioned the problem with the contact module and stated that if the problem occurred again there would be no indication to the control room. It also stated that the ASGT test relay was exercised during troubleshooting specifically to make a better connection to pass the IC-ST-RPS-0044 surveillance test. At approximately 5:06 p.m., the night shift I&C technician completed the last three steps of IC-ST-RPS-0044 with the day shift operations crew based on the data recorded by the day shift I&C technician. This consisted of ensuring the trip units were reset, removing the bypass keys, and informing the shift manager. The trip units were returned to service and an operability determination was requested by the shift manager to evaluate the ASGT test circuit during normal operation for operability. Based on discussions with I&C personnel and the shift manager, as well as review of condition reports, the inspectors questioned if the surveillance test used to declare operability had been compromised due to potential preconditioning. The inspectors also asked what corrective actions were taken to correct the problem. The inspectors brought these questions to the licensing department. These questions were documented by the licensee in Condition Report 2010-5733, on November 10, 2010. Based on discussions with licensing and I&C personnel the operating crew declared trip units A/TU-6 and A/TU-7 inoperable, replaced the contact module, performed Surveillance Test IC-ST-RPS-0044 again, and then returned the trip units to service. On November 16, 2010, operability determination associated with Condition Report 2010-5667 was completed. This determination concluded that the out of tolerance values on November 8, 2010, were not outside the design basis as the values do not account for 4 psi of margin not built into the tolerances based on Calculation FC05733. Therefore, the values could be out of tolerance +/- 16-millivolt dc before they are outside of their design basis. In addition, the increase in voltage does not affect trip unit A/TU-7 as the voltage is added to each signal, which are then subtracted to determine a difference. To address the concern regarding the ASGT test circuit effect on trip unit operability additional actions were required to confirm operability in the current calibration cycle. Specifically Work Order 396853 was generated to monitor the voltage of the relay contact on all channels to confirm operability. Surveillance Test IC-ST-RPS-0044 test frequency was increased for the next six weeks. On November 29, 2010, voltage at terminal 74 was elevated 39-millivolt dc, thus rendering trip unit A/TU-6 inoperable. This is documented in the operator logs as well as Condition Report 2010-6190. Trip Unit A/TU-6 was declared inoperable. Subsequent trouble shooting determined a bad wire in the circuit. The wire was replaced, post maintenance testing was performed, and the trip unit was returned to service. Condition Reports 200302822 and 2009-2317 documents prior out of tolerance readings, for the same values in Surveillance test IC-ST-RPS-0044, which rendered the trip unit inoperable. These events were not determined by the licensee to be functional failures. After reviewing the condition reports, the inspectors believe these particular events to be functional failures of trip unit A/TU-6. Not fixing a condition adverse to quality is a performance deficiency. The events on November 8, 2010, show that the corrective actions taken in response to Condition Report 2009-2317 were inadequate. These actions consisted of replacing the AI-31A-AW12 B2 contact module. These same actions were taken in response to Condition Report 200302822 and therefore, were within the licensees ability to foresee and correct. The actions taken in response to the events on November 8 and November 10, 2010, were inadequate as demonstrated when trip unit A/TU-6 was declared inoperable on November 29, 2010, documented in Condition Report 2010-6190. To determine if there is more than one performance deficiency, the inspectors intend to investigate the actions taken on November 8 and November 10, 2010, as well as review the licensees apparent cause analysis regarding Condition Report 2010-6190. In accordance with the guidance in Inspection Manual Chapter 0612, an unresolved item is warranted if more information is required to determine if the performance deficiency is more than minor. URI 05000285/2010005-04, Calibration Failures of RPS Trip Units 6 and 7.
05000285/FIN-2010006-01Failure to Correct Repeated Tripping of the Turbine-driven Auxiliary Feedwater Pump FW-102010Q2A self-revealing noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, occurred for the licensees failure to assure that a condition adverse to quality was corrected. Specifically, five instances were identified where the licensee failed to correct an adverse configuration design which allowed the turbine-driven auxiliary feedwater pump FW-10 exhaust backpressure trip reset lever to be bumped and unlatched which would have prevented the pump from starting when required. The failure to correct this adverse condition resulted in the turbine-driven auxiliary feedwater pump reset lever becoming unlatched and causing the pump to trip off during a urveillance test start attempt on February 17, 2010. The licensee entered this issue in their corrective action program as Condition Report CR-2010-0813. The finding is more than minor 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. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the issue screened as potentially risk significant since the finding represented a loss of system safety function of a single train for greater than the technical specification allowed outage time. The finding required a Phase 2 analysis. When evaluated per Manual Chapter 0609, Appendix A, \"Determining the Significance of Reactor Inspection Findings for At-Power Situations,\" and the Fort Calhoun Phase 2 presolved table item, Turbine-driven Auxiliary Feedwater Pump Fails to Start, the inspectors determined this finding to be potentially risk significant. A Phase 3 analysis was performed and it was determined that the finding was of very low risk significance. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensees periodic trends and assessments did not recognize the significance of precursor events related to bumping the reset lever and prompt action to prevent further problems with the turbine-driven auxiliary feedwater pump FW-10.
05000285/FIN-2010006-02Failure to Verify that the Turbine-driven Auxiliary Feedwater Pump Exhaust Backpressure Trip Lever was Fully Latched2010Q2The team identified a noncited violation of Technical Specification 5.8.1.a regarding the licensees failure to implement written procedures as recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Paragraph 3.l of Appendix A requires procedures for startup, shutdown and operation of the auxiliary feedwater system. Specifically, the licensee had no procedural guidance to verify full engagement of the turbine-driven auxiliary feedwater pump FW-10 exhaust backpressure trip mechanism when latched. This resulted in the licensees failure to identify the partially latched condition of the exhaust trip mechanism which subsequently vibrated loose during a surveillance test causing a start failure of the pump, on February 17, 2010. The licensee entered this deficiency in their corrective action program as condition Report CR 2010-0813. This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events. The finding has a crosscutting aspect in the area of problem identification and resolution associated with operating experience because the licensee failed to implement and institutionalize operating experience through changes to station operating procedures when they failed to incorporate industry information to verify the turbine-driven auxiliary feedwater pump is fully latched.
05000285/FIN-2010006-03Failure to Vent Control Oil Following Maintenance Results in Failure of the Turbine-driven Auxiliary Feedwater Pump to Start2010Q2A self-revealing noncited violation of Technical Specification 5.8.1.a was identified regarding the licensees failure to implement and maintain th applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Paragraph 9.a of Appendix A requires that suc maintenance that can affect the performance of safety-related equipment be properly preplanned and performed in accordance with documented instructions. Specifically, the licensee failed to have an adequate procedure for ensuring air was vented from the auxiliary feedwater pump control oil system following maintenance. As a result, the turbine-driven auxiliary feedwater pump failed to start during the February 26, 2009, operability test. The licensee has entered this issue into their corrective action program as Condition Report CR-2009-0905. The finding is more than minor because it is associated with the Mitigating System Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding was found to have very low safety significance (Green) because it was not a design deficiency; did not represent loss of a safety function, loss of a single train for greater than its allowed outage time, or loss of a non-technical specification required train of equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the area of human performance associated with conservative assumptions due to the licensee failing to identify possible unintended consequences of high points in a control oil system tubing design change that could become air bound and interfere with fast starts of the turbine-driven auxiliary feedwater pump.
05000285/FIN-2010006-04Turbine-driven Auxiliary Feedwater Pump Trip Due to Inadequate Design Margin2010Q2A self revealing noncited violation of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, occurred when the licensee failed to ensure that the design basis of certain structures, systems and components were translated into specifications, drawings, procedures, and instructions when implementing Engineering Change 45105. Specifically, this design change reduced the turbine-driven auxiliary feedwater pumps margin between the pump discharge pressure and the pumps high discharge pressure trip set-point resulting in an April 6, 2009, high pump discharge pressure trip during a scheduled surveillance test start. The licensee entered this issue in their corrective action program as Condition Report CR-2009-1611. The inspectors determined the finding is more than minor because it is 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. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the issue screened as potentially risk significant since the finding represented a loss of system safety function of a single train for greater than the technical specification allowed outage time. The finding required a Phase 2 analysis. When evaluated per Manual Chapter 0609, Appendix A, \"Determining the Significance of Reactor Inspection Findings for At-Power Situations,\" and the Fort Calhoun Phase 2 presolved table item, Turbine-driven Auxiliary Feedwater Pump Fails to Start, the inspectors determined this finding to be potentially risk significant. The finding was forwarded to a senior reactor analyst for review. A Phase 3 analysis was performed and it was determined that the finding was of very low risk significance. The finding has a crosscutting aspect in the area of human performance because the licensee failed to use conservative assumptions in decision making when a nonconservative design margin was approved and implemented on the turbine-driven auxiliary feedwater pump.
05000285/FIN-2010007-01Failure to Maintain External Flood Procedures2010Q2The inspectors identified an apparent violation of Technical Specification 5.8.1.a, Procedures, or failure to establish and maintain procedures that protect the intake structure and auxiliary building during external flooding events. The inspectors determined that the procedural guidance of GMRR- AE-1002,Flood Control Preparedness for Sandbagging, as inadequate because stacking and draping sandbags at a height of four feet over the top of floodgates would be insufficient to protect the vital facilities to 1014 feet mean sea level, as described in the Updated Safety Analysis Report and station procedures. The licensee has entered this condition into their corrective action program as Condition Report 2010-2387. As result of this violation, the licensee has implemented a corrective action plan to correct identified deficiencies and ensure site readiness. This performance deficiency is more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of external events and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding resulted in the degradation of equipment and functions specifically designed to mitigate a flooding initiating event. In addition, an external flood event would degrade two or more trains of a multi-train safety system. Therefore, the finding was potentially risk significant to flood initiators and a Phase 3 analysis was required. The preliminary change in core damage frequency was calculated to be 3.1 E-5/year indicating that the finding was of substantial safety significance (Yellow). The finding was determined to have a crosscutting aspect in the area of problem identification and resolution, corrective action program, for failure to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and compleXity. Specifically, from 2003 to 2008, the licensee failed to initiate appropriate corrective actions to ensure regulatory compiiance of the external flooding design basis was maintained. (P.1 (d)) (Section 40A5.1)
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05000285/FIN-2011002-01Inadequate Operating Instruction Results in a Loss of Auxiliary Feedwater2011Q1A self-revealing Green noncited violation of Fort Calhoun Station Technical Specification 5.8.1 occurred for an inadequate procedure for securing auxiliary feedwater flow when feeding the steam generators through the auxiliary feedwater ring. This inadequacy resulted in a complete loss of auxiliary feedwater for approximately three minutes. This was entered into the licensees corrective action program as Condition Report 2011-0839. The inspectors determined that the licensees inadequate operating instruction procedure was a performance deficiency. This finding was more than minor because it adversely impacted the human performance attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed the initial significance determination for the inoperable auxiliary feedwater system. The turbine-driven and motor-driven auxiliary feedwater pumps were inoperable for approximately three minutes, while the pump discharge lines were isolated during startup. The non-safety diesel-driven auxiliary feedwater pump remained available. The inspectors used the Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved an actual loss of safety function in the Mitigating Systems Cornerstone. A Region IV senior reactor analyst performed a Phase 2 significance determination and attempted to use the pre-solved worksheet from the Risk Informed Inspection Notebook for Fort Calhoun Station, Revision 2.01a. However, the pre-solved worksheet did not include the simultaneous failure of two auxiliary feedwater pumps. Therefore, the analyst performed a bounding Phase 3 significance determination. The analyst used the Fort Calhoun Standardized Plant Analysis Risk model, Revision 8.15, dated August 27, 2010, to calculate the conditional core damage probability, for a bounding event that included the failure to start for both the motor and turbine-driven auxiliary feedwater pumps. The change in core damage frequency was approximately 8.6x10-9/year. This finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
05000285/FIN-2011002-02Failure to Determine the Cause of the Out Of Tolerance Condition Regarding Reactor Protection System2011Q1The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, which states in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. Contrary to the this, between July 28, 2003, and November 29, 2010, the licensee failed to determine the cause of the out of tolerance condition impacting reactor protection system channel A trip unit 6, which was a significant condition adverse to quality. This was entered into the licensees corrective action program as Condition Report 2010-6190. The licensees repeated failure to preclude the out-of-tolerance condition regarding reactor protection system channel A trip unit 6 is a performance deficiency. This finding is more than minor because if left uncorrected, the finding could have become more significant, in that, the licensee could fall below the technical specification Minimum Operable Channels if two additional trip unit six channels (B, C, or D) became inoperable. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has a very low safety significance (Green) because it was not a design or qualification deficiency, does not represent an actual loss of safety function, nor did it screen as potentially risk significant for external events. The finding was indicative of present performance and had a crosscutting aspect in the area of human performance associated with decision-making in that the licensee failed to use conservative assumptions in decision-making. The failure of the licensee to preclude repetition of the out-of-tolerance condition of reactor protection system channel A trip unit 6 is a significant condition adverse to quality. (H.1(b))
05000285/FIN-2011002-03Failure to Submit a Timely Licensee Event Report2011Q1The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73 (a)(2)(i)(B) for the licensees failure to submit a licensee event report within 60 days of discovery. On November 29, 2010, the licensee had the available information to determine reactor protection system channel A trip unit 6 had been inoperable from November 8 until November 29, 2010. Per the licensees technical specifications, reactor protection system channel A trip unit 6 should have been in the tripped condition within 48 hours from time of discovering loss of operability. This is a reportable condition required by 10 CFR 50.73 (a)(2)(i)(B) as a condition prohibited by technical specifications. This was entered into the licensees corrective action program as Condition Report 2011-2006. The inspectors determined that the licensees failure to submit a licensee event report within the required time was a performance deficiency. The licensee had the appropriate licensing basis information as well as the inspectors specific concerns regarding inadequate troubleshooting, potential preconditioning, inadequate maintenance, and operability concern; therefore the performance deficiency was within their ability to foresee and correct. The inspectors reviewed this issue in accordance with Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was potentially affected. Specifically, the NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore a finding. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for traditional enforcement only, in accordance with the NRC Enforcement Policy. This is a Severity Level IV noncited violation consistent with Sections 2.3.2 and 6.9.d of the NRC Enforcement Policy.
05000285/FIN-2011002-04Failure to Verify Design Adequacy of Refueling Water Tank Vortex Eliminator2011Q1The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as, by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, since 1998, the licensee failed to verify the adequacy of the design of the safety injection refueling water tank vortex eliminator to prevent potential air entrainment due to vortexing in safety-related pump suction piping. This finding was entered into the licensees corrective action program as Condition Reports 2007-2452 and 2011-0311. The inspectors determined that the failure to verify the adequacy of the safety injection refueling water tank vortex eliminator 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 to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee performed subsequent analysis which demonstrated that vortexing in the safety injection refueling water tank would not impact safety-related pump operation during a design basis event. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
05000285/FIN-2011003-01Failure to Adequately Design a Reactant Coolant Pump Lube Oil Collection System2011Q2The inspectors identified a noncited violation of 10 CFR Part 50, Appendix R, Section III.O for the failure to ensure an adequate seismic design of the reactor coolant pumps oil collection system. The licensee used 2-inch copper pipe with brazed joints in the lube oil collection system. The seismic analysis of the system assumed the use of ASME Section IX during the installation of the system, but no codes or standards were used by the licensee for the brazed joints. The inspectors determined that the failure to design and install an adequate oil collection system which included provisions for the drain lines to the oil collection tank was a performance deficiency. This finding had a credible impact on safety because the inadequate installation and design of the oil collection systems presented a degradation of a fire confinement component, which had a fire prevention function of not allowing an oil leak. The inspectors determined the finding was more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors, such as a fire. The inspectors reviewed Inspection Manual Chapter 0609, Appendix F, and determined the finding was of very low safety significance, because of the low degradation rating of the fire confinement category related to the as found condition of the oil collection piping, the extremely low frequency of reactor coolant pump oil leaks, minor actual reactor coolant pump oil leaks during the past operating cycle, and other area fire protection defense-in-depth features such as automatic fire detection, manual suppression capability, and safe shutdown capability from the main control room. This finding involved a legacy issue associated with a modification for original installation; therefore, there were no assigned cross-cutting aspects.
05000285/FIN-2011003-02Failure to Follow Scaffolding Procedure2011Q2The inspectors identified a noncited violation of Technical Specification 5.8.1.a for failure to follow scaffold specification and construction Procedures SO-M-35 and PED-CSS-12. This led to the licensee declaring a number of emergency core cooling components inoperable and entering technical specification 2.0.1. The inspectors determined that not following a procedure required by Technical Specification 5.8.1.a was a performance deficiency. The finding was more than minor because if left uncorrected it would have the potential to lead to a more significant safety concern. The licensee routinely failed to perform seismic evaluations of scaffolds erected near safety-related equipment not constructed in accordance with Procedures PED-CSS-12 or SO-M-35 for preconfigured seismic scaffolding. The finding was associated with the Mitigation Systems Cornerstone while the reactor was operating; therefore, Inspection Manual Chapter 0609, Attachment 4 screening checklist was used. The finding was determined to have very low safety significance because it did not involve the total loss of any safety function, and did not contribute to external event initiated core damage accident sequences. The inspectors determined the primary cause of the finding was lack of the licensees oversight of the scaffolding program. The finding had a crosscutting aspect in the area of human performance, specifically, work practices, in that, the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported.
05000285/FIN-2011003-03Failure to Follow Radiation Work Permit Procedure2011Q2Inspectors identified a noncited violation of Technical Specification 5.8.1a for the failure to follow procedural requirements to plan and carry out decontamination work in the spent fuel pool transfer canal. On January 24, 2011, decontamination work was performed in the spent fuel pool transfer canal, using Radiation Work Permit 11-3317. While planning and controlling the work, the licensee failed to follow multiple procedure steps. Specifically, the licensee did not prepare an ALARA planning worksheet as the initial step of generating the radiation work permit, did not document justification for changing the electronic dosimeter set points which were eventually determined to be inappropriate, and did not perform an ALARA briefing before the entries were made into the spent fuel pool transfer canal, which was posted as a restricted locked high radiation area. The inspectors also determined that there were aspects of the procedure that contained vague expectations, which contributed to decisions being made without using the procedure. The failure to follow a procedure was a performance deficiency. The finding was more than minor because it negatively impacted the Occupational Radiation Safety Cornerstones attribute of program and process, in that, by not following the procedure; radiological safety attributes built into the radiation work permit program were circumvented. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the violation was 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 (4) the ability to assess dose was not compromised. This deficiency had a crosscutting aspect in the area of human performance related to work practices. Specifically, the licensee did not communicate human error prevention techniques, such as, holding pre-job briefs, self- and peer- checking, and proper documentation of activities.
05000285/FIN-2011003-04Failure to Provide Procedural Guidance to Replace or Evaluate Age Degraded Components2011Q2A self-revealing noncited violation of Fort Calhoun Technical Specification 5.8.1, Procedures, occurred due to the failure of the licensee to ensure that adequate procedures were available for maintenance which was conducted on the reactor protective systems power supplies. Specifically, there was no procedural guidance to require replacement of power supplies, or an engineering justification for continued operation, once power supplies exceeded their vendor recommended life, and/or showed signs of failure and degradation. The inspectors determined that the licensees failure to provide procedural guidance to evaluate and/or replace age-degraded components was a performance deficiency. This was a result of the licensees failure to properly implement a required procedure, and was within the licensees ability to foresee and correct and could have been prevented. This performance deficiency was more than minor because it could be reasonably viewed as a precursor to a significant event, it could lead to a loss of the reactor protective system. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Attachment 4, and determined that this finding was associated with the Mitigating Systems Cornerstone, specifically the primary degraded reactivity control contributor. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. The inspectors determined that the finding represented a qualification deficiency confirmed not to result in a loss of functionality because none of the failures to date prevented a reactor protective systems channel from tripping. Therefore, in accordance with the Phase 1 screening, the finding was of very low risk significance. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the component of operating experience because the licensee failed to adequately evaluate and communicate relevant internal and external operator experience.
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