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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000313/FIN-2018011-03 | 2018Q3 | Arkansas Nuclear | Failure to Evaluate the Effects and the Suitability of Components in Containment from a Main Steam Line Break. | The team identified an unresolved item (URI) related to the containment environment that would result from a main steam line break. Specifically, for ANO Unit 1 the licensee did not analyze the containment temperature, or evaluate the suitability of components in containment for the effects of a main steam line break (MSLB) accident. The Final Safety Analysis Report states, in part, that "At the end of Cycle 19, the original once through steam generators (OTSGs) were replaced. In support of Cycle 20 operation, an evaluation of the containment pressure/temperature response with the replacement OTSGs for loss of coolant accidents (LOCA) and MSLB was performed. For the MLSB, the containment pressure response with the replacement OTSGs was bounded by the current analysis. The post-MSLB temperature response w ith the replacement OTSGs would be worse. Entergy Operations, Inc. has adopted NUREG-0458 into the AN0-1 licensing basis which recognizes that the post-MSLB atmosphere may become superheated, but the temperature spike is of such short duration that the thermal lag of any SSC inside containment will not increase significantly. Consequently, the initial temperature peak does not define operating limits on any system, structure, or component (SSC) and the long-term containment temperature (which is essentially the saturation temperature) dominates the temperature response of SSCs. Therefore, as long as the peak MSLB pressure is less than the peak pressure following a LOCA, the temperature response of SSCs will still be defined by the LOCA." The NRC issued several bulletins subsequent to the issuance of NUREG-0458. Specifically IEB-79-01, as supplemented, and NRC Order CLI 80-21 state, in part, that "The Guidelines leave open the question of what standard will be applied to replacement parts in operating plants. Unless there are sound reasons to the contrary, the 1974 standard in NUREG-0588 will apply. The Guidelines and NUREG-0588 apply progressively less strict standards to the older plants. The justification for this position was not articulated at the time the older plants were grandfathered from the provisions of Reg. Guide 1.89." The NRC issued a Safety Evaluation Report to ANO, which states, in part, "A final rule on environmental qualification of electric equipment important to safety for nuclear power plants became effective on February 22, 1983. This rule, Section 50.49 of 10 CFR 50, specifies the requirements of electrical equipment important to safety located in a harsh environment. In accordance with this rule, equipment for Arkansas Unit 1 may be qualified to the criteria specified in either the DOR Guidelines or NUREG-0588, except for replacement equipment. Replacement equipment installed subsequent to February 22, 1983 must be qualified in accordance with the provisions of 10 CFR 50.49, using the guidance of Regulatory Guide 1.89, unless there are sound reasons to the contrary." The NRC issued Information Notice 85-39 states, in part, that the "Qualification of some replacement equipment was based on previously allowed DOR guidelines that stated "equipment is considered qualified for main steam line break environmental conditions if it was qualified for a loss-of-coolant accident environment in plants with automatic spray systems not subject to disabling single component failures." This basis of qualification is not acceptable without additional justification for replacement equipment that was procured and installed after February 22, 1983." The replacement steam generators have several design differences compared to the original steam generators. Specifically, the replacement steam generators were designed with larger secondary volumes, more tubes, flow-restricting venturis, and different materials (Alloy 690 vs. Alloy 600). Because the replacement steam generators were installed in 2005 (after 10 CFR 50.49 became effective on February 22, 1983) all replacement equipment must be qualified using the guidance of NUREG-0588 or Regulatory Guide 1.89. In addition, as stated above the licensee did not analyze or quantify the containment temperature that would result from a MSLB, and instead compared the containment pressures and the mass/energy releases that would result from a MSLB using the superseded guidance of NUREG-0458. The NRC team identified that there are several parameters that could have changed with the replacement steam generators which could impact the containment response. Specifically, input parameters such as: sub-compartment analysis, net positive suction head analysis, containment volume, heat sinks, properties of materials, heat transfer coefficients, initial conditions, and possibly cooling water temperature may affect the containment temperature response. |
| 05000313/FIN-2018011-02 | 2018Q3 | Arkansas Nuclear | Failure of Both Arkansas Nuclear One Units to Establish Adequate Corrective Actions Resulting in Excessive Instances of Damaged and Broken Internals of the Emergency Feedwater Pum o Turbine Steam Admission Check Valves. | An NRC identified Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for failure to establish an adequate corrective action program and the resulting inability to correct a deficient system design which resulted in damaged and broken internals of the check valves admitting steam to the emergency feedwater turbine. |
| 05000313/FIN-2018011-01 | 2018Q3 | Arkansas Nuclear | Failure to Properly Size the Unit 1 Emergency Diesel Generator Room Ventilation Svstems | An NRC identified Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," was identified for failure to properly size the Unit 1 emergency diesel generator room ventilation systems to be capable of removing the design heat load during the most limiting design conditions while maintaining redundancy of the exhaust fans. |
| 05000382/FIN-2017008-04 | 2017Q4 | Waterford | Potential Failure to Obtain a License Amendment for Changes to Diesel Generator Surveillance Test Interval | The team identified an unresolved item for the licensees failure to perform a 10 CFR 50.59 safety evaluation and subsequently obtain a license amendment for changes to the surveillance testing frequency of the emergency diesel generators. The licensees process for changing surveillance test intervals is controlled by Technical Specification 6.5.18, Surveillance Frequency Control Program. The licensees changes to the surveillance test intervals are made in accordance with NEI 04-10, Risk Informed Method for Control of Surveillance Frequencies, Revision 1, as written in procedure EN-DC-355, Engineering Evaluation of Proposed Surveillance Test Interval Changes, Revision 2. The team reviewed the licensees changes to the surveillance test interval, as required by Technical Specification Surveillance Requirements 4.8.1.1.2.e, for emergency diesel generators. The licensee changed the surveillance test interval for the train A and B emergency diesel generators from both emergency diesel generators tested every 18 months to each emergency diesel generator tested every 36 months. The team determined that testing the emergency diesel generators once every 36 months was contrary to guidance in Regulatory Guide 1.9, Application and Testing of Safety-Related Diesel Generators in Nuclear Power Plants, Revision 4. Specifically, Section 2.3.2.3, Refueling Outage Testing, requires the capability of the overall emergency diesel generator design should be demonstrated during every refueling outage not exceeding a period of 24 months. The team determined that the licensee did not correctly evaluate the change to the surveillance interval in accordance with surveillance frequency control program change process. Specifically, the licensee did not correctly evaluate NEI 04-10, step 1, Check for Prohibitive Commitments, and step 2, Can Commitments be Changed? of the change process. The team determined that this change would require a 10 CFR 50.59 safety evaluation and subsequent license amendment because it would result in more than a minimal increase in likelihood of a malfunction of a component important to safety as previously described in the final safety analysis report. Specifically, the test interval would no longer meet the applicable acceptance standard, Regulatory Guide 1.9, to which the licensee is committed. Planned Closure Action(s): The NRC inspectors will review the final corrective actions, pending NRC resolution of applicability of 10 CFR 50.59 to the surveillance frequency control program. Licensee Action(s): Prior to this inspection, the licensee identified this 10 CFR 50.59 issue in the corrective action program because of industry operating experience. At the time of this inspection, the licensee had not completed the final corrective action and 10 CFR 50.59 activities.These corrective actions will be completed once industry guidance on the NRC resolution of applicability of 10 CFR 50.59 to the surveillance frequency control program was available. Corrective Action Reference(s): Condition Reports CR-WF3-2017-05590 and CR-WF3-2017-5602 NRC Tracking Number: 05000382/2017008-04 |
| 05000382/FIN-2017008-03 | 2017Q4 | Waterford | Two Examples of Failure to Submit and Receive Prior Authorization of Alternatives to ASME OM Code Leak Testing Requirements | The team identified two examples of a Severity Level IV, non-cited violation of 10 CFR 50.55a(z), for failure to submit and obtain authorization prior to implementation of multiple alternatives to leak testing requirements of the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) of Nuclear Power Plants Code. Specifically, prior to November 16, 2017, the licensee did not submit and receive prior authorization to alternative leak testing requirements for safety injection valves SI-512A and SI-602B. |
| 05000382/FIN-2017008-02 | 2017Q4 | Waterford | Failure to Meet RG 1.9 Emergency Diesel Testing Requirements during Surveillance Test Results in Missed Surveillance | The team identified a Green non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification Limiting Condition for Operation 3.8.1.1 for failure to maintain operability of two separate independent diesel generators. Specifically, on May 23, 2017, the licensee failed to verify that the train A emergency diesel generator energized all auto-connected shutdown loads through the load sequencer and operated for greater than or equal to five minutes in accordance with Technical Specification Surveillance Requirement 4.8.1.1.2. |
| 05000382/FIN-2017008-01 | 2017Q4 | Waterford | Three examples of Failure to Establish and Maintain Preventive Maintenance Procedures for Safety-Related Electrical Equipment | The team identified three examples of a Green non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a, for failure to establish, implement, and maintain written procedures for activities referenced in Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, prior to November 16, 2017, the licensee failed to establish and maintain procedures covered in Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance, to implement maintenance for safety-related 1600 A, 600 V non-segregated metal-enclosed bus ducts, safety-related 4.16 kV G.E. Magne-Blast circuit breakers, and safety-related 480 V G.E. switchgear AKR breakers. |
| 05000269/FIN-2017001-03 | 2017Q1 | Oconee | Licensee-Identified Violation | Oconee Nuclear Station Technical Specification 3.0.4 requires that when a limiting condition of operation is not met, entry into a mode or other specified condition in the applicability shall not be made except when the associated actions to be entered permit continued operation in the mode or other specified condition in the applicability for an unlimited period of time. Oconee Nuclear Station Technical Specification 3.3.7, Engineered Safeguards Protective System (ESPS) Automatic Actuation Output Logic Channels, requires eight ESPS automatic actuation output logic channels to be operable in Modes 1 and 2 and Modes 3 and 4 when associated ES equipment is required to the operable. Contrary to the above, Oconee Nuclear Station Unit 1 entered Mode 4 on November 24, 2016 with ES protective system voters 1 and 2 in an abnormal configuration (bypassed) for the plant mode of operation. Operations shift personnel discovered this abnormal configuration on November 25, 2016 and restored voters 1 and 2 to an operate condition which met Technical Specification 3.3.7. This failure to maintain ESPS channels in the correct mode of operation for the required mode of applicability was a performance deficiency and was determined to be more than minor. The issue is more than minor because it was associated with the configuration control attribute of the mitigating system cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the issue challenged the configuration control attribute of ensuring operating equipment was available to respond to initiating events. The inspectors used IMC 0609, Att. 4, Initial Characterization of Findings, issued October 07, 2016, and IMC 0609, Appendix A, Significance Determination Process for Findings at Power, issued June 19, 2012, and determined the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time. The licensee has entered this issue into their corrective action program as NCR 02081523. |
| 05000269/FIN-2017001-02 | 2017Q1 | Oconee | Licensee-Identified Violation | Technical Specification 3.3.8, PAM Instrumentation, requires CHRRMs, RIA-57 and RIA-58, to be operable in Modes 1, 2, or 3. Contrary to the above, from 1998 to October 2016, the licensee failed to maintain operability of the CHRRMs for all three units when they failed to provide reasonable assurance that the CHRRMs would provide accurate measurement of containment radiation levels during a HELB event in the east penetration room of the affected unit(s). The CHRRMs are utilized in the Oconee site emergency plan and implementing procedures to support assessment of the severity of an accident. The performance deficiency was determined to be more than minor because it was associated with the facilities and equipment attribute of the emergency preparedness cornerstone and adversely affected the cornerstone objective to ensure the licensees capability to implement adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors used IMC 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, issued September 22, 2015, and determined the finding was of very low safety significance (Green) because no planning standard function failure occurred due to the availability of other parameters that could be used to validate the indications from the CHRRMs. The licensee has entered this issue into their corrective action program as NCRs 02069527 and 02077587. |
| 05000269/FIN-2017001-01 | 2017Q1 | Oconee | Failure to Comply with 10 CFR 55.49 | Green: A Green NRC-identified non-cited violation (NCV) of 10 CFR 55.49, Integrity of Examinations and Tests, was identified because the licensee engaged in an activity that compromised the integrity of examinations. Specifically, the licensee failed to ensure that current week simulator scenarios could not be predicted based on the previous weeks simulator scenarios during the annual operating exams required by 10 CFR 55.59, Requalification. While inspecting the annual operating examination schedules for the required simulator examinations for 2016 and 2017, the inspectors identified that one of the two scenarios that were administered during a single week of the annual exam cycle could be predicted for administration the following week. The licensee did not implement any immediate corrective actions because the exams were completed and there was no evidence of compromise. The licensee documented the issue in nuclear condition report (NCR) 2114313. This performance deficiency was more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, using predictable exam development and administration techniques adversely affected the integrity of the administration of the operating exams, which test licensed operator performance in order to ensure timely and correct mitigating actions during an event. Using the Licensed Operator Requalification Significance Determination Process, this finding was determined to be of very low safety significance (Green) because no known compromise of the examinations occurred. The inspectors determined the finding had a cross-cutting aspect of resources in the cross-cutting area of human performance because the licensee failed to ensure that adequate training procedures were available to meet industry standards and ensure that the potential for the compromise of regulatory examinations did not exist. (H.1) |
| 05000390/FIN-2016004-02 | 2016Q4 | Watts Bar | Inadequate Immediate Determination of Operability for Containment Penetration X-65 | Green: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to address all the design criteria for check valve, 1-CHV-31-3407, in the basis of the immediate determination of operability (IDO) for containment penetration X-65 to conclude that a reasonable expectation of operability existed. On September 19, Technical Specification (TS) compliance was restored when Penetration X-65 returned to operable when it was isolated and drained. The violation was entered into the licensees corrective action program as condition report (CR) 1216892. The performance deficiency was more than minor because it adversely affected the design control attribute of the barrier integrity system cornerstone. Specifically, reasonable assurance of operability did not exist for containment penetration X-65 from September 18, 2016, until September 19, 2016. The inspectors performed an initial screening of the finding and determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components; and hydrogen igniters are not applicable. The cause of this finding had a cross-cutting aspect of Evaluation in the area of Problem Identification and Resolution, because the licensee did not consider all functions of check valve 1-CKV-31-3407 when performing the IDO after the valve failed to pass the surveillance instruction. (P.2). |
| 05000390/FIN-2016004-03 | 2016Q4 | Watts Bar | Notice of Enforcement Discretion 16-2-01 for Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications | (Opened) Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications and Notice of Enforcement Discretion 16-2-01 Introduction: The inspectors opened an unresolved item associated with a potential noncompliance with TS 3.8.1 that occurred on October 15, 2016. Notice of Enforcement Discretion 16-2-01 was granted by the NRC staff agreeing not to enforce compliance with the TS completion time for an additional 130 hours. Description: At 6:32 a.m. on October 12, 2016, Watts Bar operations staff declared the 1A-A EDG inoperable when the output breaker to the 1A shutdown board opened unexpectedly due to phase overcurrent during performance of the load test required by procedure 0-SI-82-13, 24 Hour Load Run - DG 1A-A. The 1A-A emergency diesel generator was operating normally prior to the opening of the breaker. The licensees initial assessment determined the likely cause of the breaker trip was operation of the tap changer associated with the offsite power supply transformer. A subsequent 24 hour EDG load test was started at 12:35 a.m. on October 13, 2016. At 6:45 p.m. on October 13, 2016, operations staff noted mega volt amps (reactive) swings. During subsequent troubleshooting activities, it was determined that the mega volt amps (reactive) variance could be consistently reproduced by slight movement of a potentiometer on the 1A-A EDG voltage regulator. The licensee determined that an issue in the voltage regulator circuit was the most likely cause of the output breaker trip, and made preparations to replace and calibrate the voltage regulator on which the potentiometer was located. The licensee determined that it would require more than 72 hours to complete the removal and replacement of the voltage regulator and post-maintenance testing. The licensee requested a notice of enforcement discretion and an additional 144 hours to restore EDG 1A-A. A notice of enforcement discretion for an additional 130 hours was granted by the NRC staff at 9:30 p.m. on October 14, 2016. Consistent with NRC policy, the NRC agreed not to enforce compliance with the specific TSs in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine if there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item. (URI 05000390, 391/2016004-03, Notice of Enforcement Discretion 16-2-01 for Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications) This activity constitutes completion of one event follow-up sample, as defined in IP 71153 |
| 05000390/FIN-2016004-01 | 2016Q4 | Watts Bar | Inadequate Immediate Determination of Operability for Essential Raw Cooling Water Pumps | Green: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to base an immediate determination of operability (IDO) for essential raw cooling water (ERCW) pumps on information sufficient to conclude that a reasonable expectation of operability existed. The licensee restored compliance on November 30, 2016, when they documented an IDO that met the requirements of OPDP-8. The violation was entered into the licensees CAP as CR 1237178. The performance deficiency was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone. Specifically, reasonable assurance of operability did not exist for the ERCW pumps from November 29, 2016 until November 30, 2016. The inspectors determined the finding was of very low safety significance (Green) because it did not represent an actual loss of function for at least a single train for longer than its technical specification allowed outage time. The cause of this finding had a cross cutting aspect of Teamwork in the Human Performance area, because individuals and work groups failed to communicate and coordinate their activities within and across organizational boundaries such that nuclear safety is the overriding priority. (H.4). |
| 05000390/FIN-2016501-01 | 2016Q4 | Watts Bar | Failure to Maintain Minimum On-Shift Emergency Response Staffing Levels | Green: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50.47(b)(2) for the licensees failure to maintain the effectiveness of its emergency plan, when on more than one occasion, the number of control room operators fell below minimum staffing, as required by Appendix C of NP-REP Tennessee Valley Authority (TVA) Nuclear Power Radiological Emergency Plan (E-Plan). The licensees corrective actions included entering the issue into their corrective action program as CR 1233650. The performance deficiency was more than minor because it was associated with the emergency response organization readiness attribute of the Emergency Preparedness cornerstone and adversely impacted the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the finding in accordance with Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and using Table 5.2-1 Significance Examples for 50.47(b)(2), determined that this finding represented an example of a staffing process that would permit a shift to go below E-Plan minimum staffing requirements. The inspectors determined that the licensees process, on more than one occasion, failed to ensure that on-shift staffing met E-Plan minimum staffing requirements between March 20 and May 6, 2016. The cause of the finding was determined to be associated with the cross-cutting aspect of thorough evaluation of problems in the corrective action component of the Problem Identification and Resolution area because the organization failed to periodically analyze information from the corrective action program and other assessments in the aggregate to identify programmatic and common cause issues (P.4). |
| 05000390/FIN-2016002-10 | 2016Q2 | Watts Bar | Untimely 10 CFR 50.73 Notification of an Inoperable Rod Position Indication | The NRC identified a SL IV NCV of 10 CFR 50.73(a)(2)(i)(B) for the licensee's failure notify the NRC that the TS LCO 3.1.8 required action and completion time were not met when the analog rod position indication (ARPI) and the demand position indication system were not operable. Subsequently, the licensee submitted LER 2016-007-00 for this issue on June 20, 2016. This violation was placed in the licensees corrective action program as CR 1163150. Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this performance deficiency was dispositioned under traditional enforcement and the violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy. Using the example listed in Section 6.9.d.9, A licensee fails to make report required by 10 CFR 50.73, the issue was determined to be a SL IV violation. In accordance with IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, traditional enforcement violations are not assessed for cross-cutting aspects. |
| 05000390/FIN-2016002-06 | 2016Q2 | Watts Bar | Failure to Satisfy TS LCO 3.6.3 | The NRC identified a Green NCV of TS for the failure to recognize and take the required actions in TS 3.6.3 for inoperable containment penetration flow paths. Specifically, the required actions of TS 3.6.3 applied on November 21, 2015, and were not taken until January 30, 2016. Upon discovery, on January 30, 2016, the affected containment penetrations were isolated by placement of a clearance, thereby satisfying the TS required actions. The licensee entered the violation into the CAP as CR 1172114. The performance deficiency was more than minor because the ERCW supply and discharge containment penetrations for the 1D upper containment cooler were inoperable for longer than the TS allowed outage time. Because the 1D upper containment cooler ERCW containment penetrations were inoperable and resulted in the failure to satisfy TS LCO 3.6.3, reasonable assurance of the integrity of the containment design barrier was adversely affected. The inspectors determined the finding was of low safety significance (Green) because the upper containment cooler ERCW penetrations are small lines (<1-2 inches in diameter) and IMC 0609, Appendix H Containment Integrity Significance Determination Process dated May 6, 2004, Table 4.1 states that small lines (<1-2 inches in diameter) would not generally contribute to LERF. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee failed to make the prudent choice to fully evaluate the unsuccessful surveillance test on November 15, 2015, and instead simply documented the issue in the corrective action program and deferred the solution, resulting in the TS violation six days later. |
| 05000390/FIN-2016002-02 | 2016Q2 | Watts Bar | Failure to Translate Design Requirements into a Maintenance Procedure for the 1B-B Charging Pump Room Cooler | The NRC identified a Green NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control for the licensees failure to specify nominal shaft size along with specific acceptance criteria for shaft tolerance measurements for the 1B-B centrifugal charging pump (CCP) room cooler fan shaft. The licensee repaired the room cooler by replacing the fan shaft and the finding was entered into the licensees corrective action program as CR 1146474. The performance deficiency was more than minor because it affected the equipment performance attribute of the mitigating system cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors determined that this finding required a detailed risk analysis since it represented an actual loss of function of a single train for greater than its TS-allowed outage time. The finding does not present an immediate safety concern because the licensee has verified current operability. A Senior Reactor Analyst evaluated the increase in core damage frequency due to the pump being non-functional over the exposure period and determined it was 3.6E-7/year (Green). The dominant scenario was a loss of component cooling water, which combined with a loss of RCP seal injection causes a loss of coolant accident and leads to core damage. The risk increase was very low because of the limited exposure time, the availability of the opposite train pump, and the time dependent nature of the pump failing due to lack of room cooling. The inspectors determined that the finding had a cross-cutting aspect of design margin in the area of Human Performance because the licensee failed to carefully guard margins through a systematic and rigorous process. Specifically, the translation of shaft diameter from design documents into 0-MI-0.16 lacked rigor and allowed an undersized shaft to go undetected, leading to cooler failure. |
| 05000390/FIN-2016002-01 | 2016Q2 | Watts Bar | Failure to Ensure that a Train of Source Range Detection was Available to Monitor Neutron Population During a Fire Event | The NRC identified a Green NCV of Operating License Condition 2.F for the licensees failure to ensure that a train of source range detection was available to monitor neutron population during the initial stages of a fire event on Unit 1. This issue was entered into the licensees corrective action program as CR 1098240. The licensees failure to ensure a train of source range detection was free from fire damage was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to maintain the capability to monitor neutron population during the early stage of a fire event. In accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, the finding was determined to be of very low safety significance (Green) because the reactor would have been able to reach and maintain a stable plant condition. No cross-cutting aspect was identified for this issue. |
| 05000390/FIN-2016002-09 | 2016Q2 | Watts Bar | Untimely 10 CFR 50.73 Notification of Failure to Meet Technical Specification Surveillance Requirement 3.5.2.3 for the Emergency Core Cooling System | The NRC identified a SL IV NCV of 10 CFR 50.73(a)(2)(i)(B) for the licensee's failure to report, within 60 days of discovery, a condition which was prohibited by the plants TS associated with recent performances of TS surveillance requirement (SR) 3.5.2.3 for verification that emergency core cooling system (ECCS) piping is full of water. Subsequently, the licensee submitted LER 2016-003-00 for this issue on May 10, 2016. This violation was placed in the licensees corrective action program as CR 1166564. Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this performance deficiency was dispositioned under traditional enforcement and the violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy. Using the example listed in Section 6.9.d.9, A licensee fails to make report required by 10 CFR 50.73, the issue was determined to be a SL IV violation. In accordance with IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, traditional enforcement violations are not assessed for cross-cutting aspects. |
| 05000391/FIN-2016002-08 | 2016Q2 | Watts Bar | Failure to Follow Maintenance Procedure Results in overspeed trip of the 2C-S Turbine Driven Auxiliary Feedwater Pump | A self-revealed Severity Level (SL) IV non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified at Watts Bar Unit 2 for the licensees failure to follow procedure 0-MI-1.003, Disassembly, Inspection, and Reassembly of Auxiliary Feedwater Pump Turbine. Specifically, the valve stem spring coil gap was not set in accordance with procedure, causing the turbine-driven auxiliary feedwater (TDAFW) pump to trip on electrical overspeed when the level control valves (LCVs) were closed. This issue was corrected on May 30, 2016, when the proper spring coil gap was set and verified and the post maintenance test was performed satisfactorily. The issue was entered into the licensees corrective action program as CR 1175968. The performance deficiency was more than minor because it represented an improper or uncontrolled work practice that could impact quality or safety involving safety-related structures, systems, and components (SSCs). The finding was a SL IV violation because it represented a failure to meet a regulatory requirement, specifically a quality assurance (QA) criteria to follow quality-related procedures, which had more than minor safety significance. The finding was assigned a crosscutting aspect of resources in the Human Performance area because the licensee failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the procedure that set the coil spring gap lacked sufficient detail and rigor to ensure that the coil gap would be set appropriately by the technicians. |
| 05000390/FIN-2016002-07 | 2016Q2 | Watts Bar | Untimely 10 CFR 50.73 Notification of Inoperable Containment Penetrations | The NRC identified a SL IV NCV of 10 CFR 50.73(a)(2)(i)(B) for the licensee's failure notify the NRC that the TS LCO 3.6.3 required action and completion time were not met for an inoperable emergency raw cooling water (ERCW) containment isolation valve. Subsequently, the licensee submitted LER 2016-009-00 for this issue on July 15, 2016. This issue was placed in the licensees corrective action program as CR 1174000. Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this performance deficiency was dispositioned under traditional enforcement and the violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy. Using the example listed in Section 6.9.d.9, A licensee fails to make report required by 10 CFR 50.73, the issue was determined to be a SL IV violation. In accordance with IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, traditional enforcement violations are not assessed for cross-cutting aspects. |
| 05000391/FIN-2016002-05 | 2016Q2 | Watts Bar | Failure to Perform A TDAFW Surveillance In Accordance With Procedures | The NRC identified a SL IV NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, at Watts Bar Unit 2 for the licensees failure to follow the surveillance test program procedure by making adjustments to the turbine-driven auxiliary feedwater (TDAFW) pump control system during the performance of a surveillance instruction. The licensee reperformed the surveillance instruction with satisfactory results. The issue was entered into the licensees corrective action program as CR 1167102. The performance deficiency was more than minor because making adjustments to the TDAFW pump control system during the performance of a surveillance instruction could invalidate the test and result in the TDAFW pump being inappropriately declared operable. As described in IMC 2517, the significance of this issue was determined using traditional enforcement, because the cornerstone associated with this finding was not being assessed by the reactor oversight process (ROP). The inspectors determined this finding to be of very low safety significance, SL IV, because it represented a failure to meet a regulatory requirement, specifically a QA criteria to follow quality-related procedures, which had more than minor safety significance. The finding was assigned a cross-cutting aspect of Conservative Bias in the Human Performance area because numerous individuals were aware the speed adjustment had been made while completing the surveillance instruction but did not question the appropriateness of that adjustment until prompted by NRC inspectors. |
| 05000391/FIN-2016002-04 | 2016Q2 | Watts Bar | Failure to Follow Operability Procedure Results in Potential Inoperability of the 2A-A Auxiliary Feedwater Pump | The NRC identified a SL IV NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, at Watts Bar Unit 2 for the licensees failure to follow procedure OPDP-8, Operability Determination Process and Limiting Condition for Operation Tracking, Revision 22. Specifically, the 2A-A motor-driven auxiliary feedwater pump (MDAFW) was potentially inoperable in mode 3 due to inadequate compensatory measures that were being controlled outside of the operability process. The issue was corrected and the pump returned to operable status on April 19, 2016. The issue was entered into the licensees corrective action program as CR 1163431. The performance deficiency was more than minor because it represented an improper or uncontrolled work practice that could impact quality or safety, involving safety-related SSCs. Specifically, failure to appropriately use the operability process when measures must be established to compensate for degraded or nonconforming conditions can lead to SSC inoperability. As described in IMC 2517, the significance of this issue was determined using traditional enforcement, because the cornerstone associated with this finding was not being assessed by the reactor oversight process (ROP). The inspectors determined this finding to be of very low safety significance, SL IV because it represented a failure to meet a regulatory requirement, specifically a quality assurance (QA) criteria to follow quality-related procedures, which had more than minor safety significance. The finding was assigned a cross-cutting aspect of Work Management in the Human Performance area because the minor maintenance work order created to compensate for the oil loss from the 2A-A MDAFW pump was never reviewed by operations, which could have identified the out of process error. (H.5). |
| 05000390/FIN-2016002-03 | 2016Q2 | Watts Bar | Untimely 10 CFR 50.73 Notification of an Inoperable Charging Pump | The NRC identified a Severity Level (SL) IV non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50.73(a)(2)(i)(B) for the licensee's failure to notify the NRC that the technical specification (TS) limiting condition for operation (LCO) 3.5.2 required action and completion time were not met when the 1B-B centrifugal charging pump (CCP) was inoperable due to an inoperable room cooler. Subsequently, the licensee submitted LER 2016-006-00 for this event on June 30, 2016. This issue was placed in the licensees corrective action program (CAP) as CR 1165380. Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this performance deficiency was dispositioned under traditional enforcement and the violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy. Using the example listed in Section 6.9.d.9, A licensee fails to make report required by 10 CFR 50.73, the issue was determined to be a SL IV violation. In accordance with IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, traditional enforcement violations are not assessed for cross-cutting aspects. |
| 05000390/FIN-2016001-10 | 2016Q1 | Watts Bar | Failure to Maintain an Adequate Surveillance Procedure for Emergency Core Cooling System Venting | The inspectors identified an apparent violation of TS 5.7.1.1.a, Procedures, for the licensees failure to maintain procedure 1-SI-63-10.1-A, ECCS Discharge Pipes Venting Train A Inside Containment, Revisions 11-16, in accordance with the requirements of Regulatory Guide 1.33. Specifically, the procedure did not have provisions for quantifying accumulated gases during venting which allowed emergency core cooling system (ECCS) piping to be vented without being evaluated for potential adverse impacts on system operability. The licensee implemented manual ultrasonic testing (UT) of gas accumulation and entered this issue into their corrective action program as CR 1136359. The performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the potential existed for an unacceptable void affecting ECCS operability to develop prior to the next scheduled surveillance. The inspectors determined the finding could not be screened to GREEN and may require a detailed risk evaluation following a determination of whether the finding represents a loss of system and/or function. Because the safety characterization of this finding is not yet finalized, it is being documented with a significance of To Be Determined (TBD). The inspectors determined that the finding had a cross-cutting aspect of Change Management in the area of Human Performance because the licensee failed to use a systematic process to implement changes to the ECCS venting procedure to ensure that Generic Letter 2008-01 commitments would continue to be met. |
| 05000390/FIN-2016001-08 | 2016Q1 | Watts Bar | Charging Pump 1B-B Room Cooler Fan Bearing Failure | Inspectors identified an unresolved item (URI) associated with the failure of the 1B-B charging pump room cooler. This item is unresolved pending review of an equipment apparent cause evaluation that was performed after deficiencies were identified by inspectors in the past operability evaluation. On September 27, 2015, the licensee installed a new bearings on the 1B-B CCP room cooler fan shaft as part of planned maintenance (PM) under WO 115790759. The WO noted the room cooler had a broken lubrication line close to the point where it is attached to the outboard fan shaft bearing, but the new bearing on the fan shaft, including the outboard shaft bearing, were installed without an immediate repair of the lubrication line. The bearing replacements for WO 115790759 were accomplished in accordance with maintenance procedure 0-MI-0.16, Maintenance Guidelines for Belt Driven Equipment, Rev. 7. Appendix D, Bearing Installation, Step 14 requires, All remote lubrication lines, remote vibration attachments, etc. shall be verified as attached prior to return to service. The work order noted at this step that the lubrication line to the outboard fan shaft bearing was broken in half and will need to be replaced prior to return to service and the step was left blank. The licensee did not initiate a CR for this degraded condition. Due to the broken lubrication line, the outboard fan shaft bearing was the only fan shaft bearing that was not greased during installation. October 15, 2015, the licensee completed the PMT for the room cooler and noted it to be satisfactory. The broken lubrication line documented in the PM WO was identified and CR 1093983 was initiated to document the condition. This CR stated that the broken lubrication line did not affect the functionality of the fan and could be repaired at the next scheduled PM. This assessment was not questioned during the review of the CR for operability. The fan was returned to service and declared operable. On December 4, 2015, the room cooler failed in service. The licensee declared the 1BB charging pump inoperable and entered the applicable TS LCO. Investigation revealed that the outboard fan shaft bearing had failed. At this point, the inappropriate treatment of the degraded lubrication line under 0-MI-0.16 and the associated PMT was identified. This issue was documented in the licensees CAP in CR 1111791. The licensee performed a past operability evaluation (POE) for CR 1111791 which concluded the fan was operable until several hours before the time of the failure. The POE was based largely on statements from the bearing vendor indicating that the new bearing was pre-lubricated at the factory and should have performed under load for a long period of time without needing to be pre-greased at installation. The POE was hampered by the fact that the licensee did not retain the damaged bearing for failure analysis. The inspectors reviewed the POE and determined that it failed to adequately document sufficient information to either discount the broken lubrication line as a cause of the bearing failure or to identify another cause. In response, the licensee opened an investigation of the cause of the bearing failure under an equipment apparent cause evaluation. Because more information is necessary to evaluate the cause of the 1B-B CCP room cooler fan shaft bearing failure, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to review the equipment apparent cause evaluation, which was not completed by the end of the inspection period. This is identified as URI 05000390/2016001-08, Charging Pump 1B-B Room Cooler Fan Bearing Failure. |
| 05000390/FIN-2016001-09 | 2016Q1 | Watts Bar | Appropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant Response | The inspectors identified an URI associated with the timely and effective corrective action associated with an adverse trend in safety related pump performance, including mechanical seal degradation and failure. This item is unresolved pending review and evaluation of the licensees response to the CRs generated to determine if a performance deficiency exists. During Unit 1, 2015 fall outage, the 1A Safety Injection (SI) pump mechanical seal was replaced. The mechanical seal had degraded to a point at which the leakage was greater than the Technical Specification limit for ECCS leakage outside of containment. The inspectors identified several issues during a review of the Prompt Determination of Operability for CR 1125623 and WO 116050574 to replace the seal. Specifically, inspectors found that non-QA1 parts were being used for seal replacement, the seal was the original equipment manufacturer part from startup, the failure mechanism was not clearly understood, and an extent of condition review was not performed. The inspectors reviewed other safety related pump mechanical seal performance and corrective action program entries. The inspectors are awaiting the completion of the licensees evaluation to determine the licensees compliance with applicable procedures and TS relative to pump operability and ECCS leakage limits outside containment. Additional inspection activities are needed to determine the extent of condition and compliance with the procedures and TS. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000390/2016001-09, Appropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant Response. |
| 05000390/FIN-2015010-01 | 2015Q3 | Watts Bar | 420 Minute Operator Manual Action to Provide Source Range Monitoring Capability | The inspectors identified an unresolved item associated with a fire protection safe shutdown OMA that established a time requirement of 420 minutes to provide a functional source range monitor. The inspection team noted that procedure 1-AOI-30.2 C36, Fire Safe Shutdown Room 737-A1A, Rev. 0005 included a 420 minute operator manual action (OMA) to establish a functional source range monitor. The OMA was listed as OMA 649 in Calculation EDQ00099920090016, Appendix R Unit 1 & 2 Manual Action, Rev. 4. The inspectors also noted the following: - Westinghouse Owners Group letter, WOG-05-36 (dated 01/28/2005), Section 6.2, Long Term Cold Shutdown Capability, stated that typical instrumentation to achieve a shutdown condition during Appendix R event included the source range monitors. - Technical Specification 3.3.1.L required an operable source range neutron flux channel in Modes 3, 4, and 5; and stipulated that positive reactivity additions (such as plant cooldown) be suspended when the instrument was inoperable. - Procedure 1-AOI-30.2, Fire Safe Shutdown, Rev. 0005, Step 5.3.15, stated that at least one channel of nuclear instrumentation indication must be available to monitor shutdown neutron population. - Procedure 1-AOI-30.2 C36 included a note that stated that RCS cooldown should not be initiated until source range monitoring capability can be assured. - Procedure 1-AOI-30.2 C36 directed operators to depressurize and cooldown an action that was typically required at 60 75 minutes. The 420 minute OMA would allow shutdown and subsequent cooldown of the reactor plant without operators having the ability to monitor neutron population. The licensee contended that OMA 649 was part of the sites licensing bases and thus the capability to monitor source range was not required until 420 minutes. The inspection team determined that this issue required additional inspection because the licensee did not provide an alternative method to monitor neuron population and did not provide adequate restrictions to prevent cooldown activities until monitoring capability was restored. Additionally, the OMA conflicted with the technical specification requirements for source range availability. The issue is unresolved pending additional review to determine if a performance deficiency exists. Required actions to resolve this issue include a detailed review of applicable docketed licensing bases correspondence; consultation with NRRs fire protection and technical specification branches; and an assessment to determine the applicable fire areas if the issue is to be determined to be a more-than-minor performance deficiency. This issue will be tracked as URI 05000290/2015010-01, 420 Minute Operator Manual Action to Provide Source Range Monitoring Capability. |
| 05000413/FIN-2015301-01 | 2015Q2 | Catawba | Licensee-Identified Violation | The following Severity Level IV violation was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. Following the facilitys administration of the initial written examination on May 28, 2015, the licensee identified that an earlier version of the examination was inadvertently provided to the RO applicants. The licensee immediately informed the NRC. The earlier version of the examination did not include the changes that were made to resolve NRC comments provided during the preexamination review of the written examination. This earlier version of the examination had not been approved by the NRC for administration to the license applicants. 10 CRF 55.49, Integrity of examinations and tests states, in part, that facility licensees shall not engage in any activity that compromises the integrity of any test or examination required by this part. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination. Contrary to the above, on May 28, 2015, the licensee administered an unapproved RO written examination, an activity that compromised the integrity of the written examination. The RO applicants did not get the benefit of the question enhancements that occurred during the examination review. The SRO applicants experienced a higher quality exam than that of the RO applicants. This is neither equitable nor consistent. The unapproved version of the exam was subsequently reviewed by the NRC and was determined to be valid. See enclosure 3 to this report to review the analysis of the administered written exam for validity of the exam. A violation of 10 CFR 55.49 is a violation that potentially impacts the regulatory process, because the examination results are used by the NRC to make licensing decisions. An improperly administered examination has the potential to provide inaccurate information to the NRC regarding the competence of the applicants. There were no actual or potential safety consequences. This violation is being treated as a Severity Level IV non-cited violation consistent with Section 2.3.2.a. of the NRC Enforcement Policy. The violation was entered into the licensees corrective action program as Nuclear Condition report 01931989. |
| 05000528/FIN-2015301-01 | 2015Q2 | Palo Verde | Licensee-Identified Violation | The following violation of very low safety significance (Green) and Severity Level IV was identified by the licensee and is a violation of NRC requirements, which meets the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. Title 10 CFR 55.49, Integrity of Examinations and Tests, requires, in part, that facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Contrary to the above, on April 14, 2015, the licensee engaged in an activity that compromised the integrity of the examination. Specifically after administrative JPMs had been administered to the applicants by the examination team, the licensee, upon performing their examination security walk-down, neglected to secure a three-ring binder that contained two reactor operator and two senior reactor operator administrative JPMs that were to be performed the next day. All four JPMs were left unattended and unsecured until 5:00 a.m. on April 15, 2015, when they were discovered as part of the licensee examination security preparation procedure. The four compromised JPMs were replaced by new administrative JPMs as required by NUREG-1021. The failure to meet 10 CFR 55.49 was evaluated through the traditional enforcement process because it impacted the ability of the NRC to perform its regulatory oversight function. This resulted in assignment of a Severity Level IV violation because it involved a non-willful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. The associated performance deficiency was screened as Green because there was not an actual effect on the equitable and consistent administration of any examination required by 10 CFR 55.59, Integrity of Examinations and Tests. The licensee entered this issue into their corrective action program as PVAR 4645293. |
| 05000317/FIN-2014004-01 | 2014Q3 | Calvert Cliffs | Main Steam Line Drain Containment Isolation Valves not Scoped in In-Service Testing | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, Codes and Standards, for Exelons failure to meet the test requirements set forth in the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for main steam line drains (MSLDs) and containment isolation valves (CIVs) motor operated valves (MOVs) (6611, 6612, 6613, 6615, 6620, 6621). Specifically, Exelon failed to scope the MSLD MOVs in their in-service testing (IST) program. As a result, the MOVs reliability was not ensured due to valve degradation not being trended as required in the IST program. Also, the MOV operability was in question because the valves were never tested to perform their containment isolation function. Exelon entered this issue into their corrective action program (CAP) as condition report (CR)-2014-005961. Immediate corrective actions included testing the MOVs. The inspectors determined that the failure to scope and meet the testing requirements of the OM Code for MSLD MOVs in accordance with 10 CFR 50.55a was a performance deficiency. This finding is more than minor because it was associated with the barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to scope and test the MSLD MOVs in accordance with the OM Code did not ensure component reliability by monitoring valve degradation and did not provide assurance that the MSLD MOVs would perform their CIV function in order to protect the public from radionuclides releases during a steam generator tube rupture (SGTR) with a loss of offsite power event. The inspectors reviewed IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions issued June 19, 2012, and determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and the finding did not involve an actual reduction of hydrogen igniters in the reactor containment. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency was not reflective of current licensee performance. Specifically, the 2007 IST fourth year interval submittal was the last reasonable opportunity for Exelon to identify this issue. |
| 05000483/FIN-2014004-03 | 2014Q3 | Callaway | Licensee-Identified Violation | The licensee dose assessment methods are inaccurate under some circumstances. Title 10 of the Code of Federal Regulations, Section 50.54(q)(2) requires, in part, that power reactor licensees follow and maintain the effectiveness of an emergency plan that meets the requirements of Appendix E to Part 50 and the standards of 10 CFR 50.47(b). Section 50.47(b)(9) requires that adequate systems, methods, and equipment for assessing the actual and potential offsite consequences of a radiological emergency condition are in use. Contrary to the above, between December 5, 2009, and April 23, 2014, Callaway Plant did not maintain the effectiveness of an emergency plan that fully met the standards of 10 CFR 50.47(b). Specifically, the licensee failed to maintain adequate systems, methods, and equipment for assessing the actual and potential offsite consequences of a radiological release to the environment. The licensee identified circumstances which could cause their dose assessment program to overestimate offsite doses by a factor of 24. This finding was assessed using Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and was determined to be of very low safety significance (Green). The finding was a failure to comply with NRC requirements, was not a loss of risk significant planning standard function, and was not a degraded planning standard function. The finding was determined not to degrade the planning standard function because the calculation was not inaccurate in its normal operating configuration, the circumstances in which the inaccurate calculation would be used were rare, a method existed for the operator to correct the error, and because the error could be readily detected from examination of the dose assessment report. The issue was entered into the licensees corrective action program as Corrective Action Request 201402814. |
| 05000483/FIN-2014004-02 | 2014Q3 | Callaway | Failure to Verify Material Properties Prior to Installation | The inspectors reviewed a self-revealing finding involving failure to verify the proper material was installed in the plant during a modification to the circulating water pumps. Specifically, Request for Resolution 201300416 specified the use of ASTM A276 410 stainless steel cap screws with a tensile strength around 186 ksi. Contrary to this, 410 stainless steel cap screws with a tensile strength between 201 ksi and 221 ksi were installed. Because the tensile strength was much higher, and thus more brittle and susceptible to stress corrosion, these cap screws were not appropriate for the application. This led to failure of the cap screws and the separation of the shaft coupling for circulating water pump B after less than one operating cycle in service, degrading condenser vacuum. The licensee removed the modification and installed the original type cap screws. This issue was entered into the licensees corrective action program as Callaway Action Request 201404722. The inspectors determined that failure to verify the correct materials were installed in the plant during a modification was a performance deficiency. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as during power operations. Specifically, failure to install the correct material resulted in failure of circulating water pump B and degrading condenser vacuum. The inspectors evaluated the finding using NRC Inspection Manual 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined the finding was of very low safety significance (Green) because the transient initiator did not cause a reactor trip and the loss of mitigating equipment. This finding has an avoid complacency cross-cutting aspect within the human performance area because the licensee relied on the vendor to provide the correct material and did not verify the cap screws met the material specification. |
| 05000317/FIN-2014004-02 | 2014Q3 | Calvert Cliffs | Licensee-Identified Violation | TS 3.4.10, Pressurizer Safety Valves, requires two pressurizer safety valves to be operable during Modes 1 and 2, and in Mode 3 when all RCS cold leg temperatures are greater than 365F for Unit 1 or 301F for Unit 2. With one pressurizer safety valve inoperable, TS 3.4.10, Condition A, requires the inoperable valve to be restored within 15 minutes. If this is not able to be completed or if two pressurizer safety valves are inoperable, then TS 3.4.10, Condition B, is entered which requires the unit to be in Mode 3 within 6 hours AND the unit to be cooled down to below 365F for Unit 1 or 301F for Unit 2 within 12 hours. Contrary to the above, on March 12, 2013, Unit 2 pressurizer safety valve BNO4375, which had been installed in position 2RV200 during the previous operating cycle, was measured higher than its TS allowable value during as-found lift point testing. On February 28, 2014, Unit 1 pressurizer safety valves BN04373 and BM07952, which had been installed in positions 1RV200 and 1RV201 respectively during the previous operating cycle, were measured lower than their TS allowable value during as-found lift point testing. In both cases, the valves had been replaced with tested, operable valves prior to discovery of the as-found condition. Exelon concluded that the valve had been inoperable for a period of time greater than the allowed TS outage times specified in TS 3.4.10. Exelon entered both issues into their CAP as CR-2013-002415, CR- 2014-002236, and CR-2014-002237. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that each example was a finding of very low safety significance (Green) because the finding did not represent an actual loss of the pressurizer safety valve systems credited safety function to relieve pressure to prevent RCS pressure from exceeding 110 percent of RCS pipings design pressure. |
| 05000483/FIN-2014004-01 | 2014Q3 | Callaway | Failure to Perform Nondestructive Testing on Essential Service Water Piping in Accordance with Procedures | The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform nondestructive testing on portions of the essential service water system known to be susceptible to wall thinning in accordance with procedures. As a result, the licensee failed to identify wall thinning prior to developing a through-wall leak that rendered train A inoperable. Specifically, despite procedural guidance to the contrary, technicians only used the low frequency electromagnetic technique testing, which cannot monitor bends and portions of welds. They also failed to properly calibrate this equipment, and failed to perform ultrasonic testing on the portions of essential service water system that could not be properly monitored by use of low frequency electromagnetic technique. The resultant through-wall leaks were repaired according to ASME code standards. The licensee entered this issue into their corrective action program as Callaway Action Request 201405200 and planned to re-perform testing during the fall of 2014. Failure to follow procedures while performing nondestructive testing on portions of the essential service water system was a performance deficiency. This performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to perform nondestructive testing on portions of the essential service water system that were known to be susceptible to wall thinning resulted in the failure to prevent a through-wall leak affecting the availability of a safety related system. Using NRC Inspection Manual 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance because it only affected a single train, and resulted in a loss of function for less than its technical specifications allowed outage time. This finding has a procedure adherence cross-cutting aspect within the human performance area because the licensee failed to ensure that individuals followed processes, procedures, and work instructions. Specifically, licensee oversight failed to ensure that contractors followed specific guidance in their procedures for both ensuring that the low frequency electromagnetic technique tool was appropriately calibrated and areas unable to be scanned were tested utilizing ultrasonic testing. |
| 05000317/FIN-2014003-01 | 2014Q2 | Calvert Cliffs | Main Steam Line Drain Containment Isolation Valves not Scoped in IST | An unresolved item (URI) was identified by the inspectors relating to an issue regarding the failure of Exelon to scope main steam line drains (MSLDs) and CIVs motor operated valves (MOVs) (6611, 6612, 6613, 6615, 6620, and 6621) into their inservice testing (IST) program. Description: The inspectors identified an issue of concern involving Exelons scoping of MSLD MOVs into the IST program. The MSLD MOVs are normally open valves with the ability to be remotely-operated from the main control room. The MSLD MOVs are classified as CIVs per UFSAR, Figure 5-10, Containment Structure Isolation Valve Arrangement, Sheet 24 and 25. This figure classifies the main steam penetrations as Type III, and requires the valves to be closed to perform their CIV function. UFSAR, Section 5.2, Isolation System, Subsection 5.2.2, System Design, defines a Type III penetration as a line not directly connected to the reactor coolant system (RCS) or the containment structure atmosphere that has at least one valve, either a check valve or a remotely-operated valve, outside of the containment structure. These valves are classified as American Society of Mechanical Engineers (ASME) Code, Class 2, per drawing 60740, Sheet 0001, Steam Line Drainage System, Revision 39, and M-601, Piping Class Summary Sheets, Revision 49. The ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2004, Subsection ISTA, General Requirements, Section ISTA-1100, Scope, states in part, Section IST establishes the requirements for pre-service and IST and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. These requirements apply to: a) pumps and valves that are required to perform a specific function in shutting down the reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident. 10 CFR 50.55a(f)(1), Codes and Standards, requires the establishment of OM Code IST test requirements to components which are classified ASME Code Class 1, 2 and 3. The inspectors require additional information from Exelon to determine if there is a performance deficiency which is more than minor. Specifically, the revision to calculation CA06453, Steam Generator Tube Rupture Accident Using Source Terms; calculation referenced in April 6, 1988, memo from D. S. Elkins to B. B. Mrowca, Impact of the Main Steam Drain Line on the 10CFR100 Limits of the Steam Generator Event; and for CCNPP to research which standard for the design of CIVs the plant was licensed to (equivalent to ANSI N271-1976, Containment Isolation Provisions for Fluid Systems.) The issue is identified as (URI 05000317/318/2014003-01, Main Steam Line Drain Containment Isolation Valves not Scoped in In-Service Testing Program.) |
| 05000250/FIN-2014404-04 | 2014Q2 | Turkey Point | Licensee-Identified Violation | |
| 05000250/FIN-2014404-03 | 2014Q2 | Turkey Point | Licensee-Identified Violation | |
| 05000250/FIN-2014404-02 | 2014Q2 | Turkey Point | Licensee-Identified Violation | |
| 05000251/FIN-2014404-01 | 2014Q2 | Turkey Point | Licensee-Identified Violation | |
| 05000482/FIN-2014002-01 | 2014Q1 | Wolf Creek | Inadequate Work Instructions for Reinstallation of ESW Expansion Joints | The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, Procedures, for maintenance instructions inappropriate to the circumstances. Specifically, Work Orders 11-341986-005 and 11-342065-002 did not contain adequate instructions for reassembling essential service water Garlock expansion joints to ensure proper joint alignment. As a result, on February 11, 2014, the inspectors identified that the inlet expansion joint for the essential service water intercooler heat exchanger, which provides cooling to emergency diesel generator B jacket water system, was misaligned by 0.5 inches, which exceeded the vendor specification of less than 0.125 inch. This item was entered into the corrective action program as Condition Reports 79352 and 79623, and the fitting was replaced during the mid-cycle 2014 outage. The licensee also conducted an extent of condition inspection and identified three additional Garlock expansion joints that were not made with the approved liner material. The failure to properly reinstall essential service water expansion joints consistent with the vendor approved and analyzed configuration was a performance deficiency. The performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the misaligned Garlock expansion joint in the essential service water system degraded its long-term operability and its ability to withstand a seismic event. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time and the finding did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program for greater than 24 hours. Specifically, although the expansion joint was in a degraded condition, it was determined to be operable based on an engineering evaluation and seismic test data. The inspectors determined that the finding had a cross-cutting aspect in the human performance area of resources because the licensee did not ensure that personnel equipment, procedures, and other resources were available and adequate to support nuclear safety. |
| 05000482/FIN-2014002-03 | 2014Q1 | Wolf Creek | Failure to Maintain Seismic and Missile Protection Design Basis Requirements During Essential Service Water Construction | A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the failure to conduct excavation work such that it would ensure that design basis requirements for tornado missile protection and seismic qualification of safety-related cables were maintained during construction near the essential service water pump house. Specifically, when excavation near underground essential service water cables caused a loss of safety-related backfill over the cables, the licensee did not plan and execute the work in a manner that ensured that the qualified soil coverage around the train B essential service water duct bank was maintained by protecting against trench cave-ins. Failure to maintain adequate soil coverage of the essential service water duct banks during construction is a performance deficiency. The deficiency is more than minor because it affected the protection against external factors and design control attributes of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Events Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. The inspectors determined that the finding had a cross-cutting aspect of work management in the area of human performance in that the process for planning, controlling, and executing work did not adequately include the identification and management of risk. Specifically, work planning did not account for adequate shoring material to prevent design basis ground cover from caving in during planned excavations in the vicinity of operable safety related equipment. |
| 05000482/FIN-2014002-04 | 2014Q1 | Wolf Creek | Licensee-Identified Violation | A licensee-identified violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for failure to ensure that design basis requirements were maintained in response to a cave-in of required essential service water ground cover. Specifically, the licensee did not ensure that the qualified soil coverage around the train B essential service water duct bank was maintained when they re-covered the duct banks with an unapproved material. Contrary to these requirements, on January 20, 2014, upon a loss of essential service water duct bank soil coverage due to a cave-in, the licensee refilled the voided area with an unapproved material that was not qualified to withstand seismic and missile design basis accidents. The performance deficiency was failure to ensure the appropriateness of seismic and missile-qualified material. This violation was more than minor because it affected the protection against external factors and design control attributes of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Events Screening Questions, the inspectors determined that the finding was of very low safety significance because it did not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. Since the finding was licensee-identified, no cross-cutting aspect is assessed. The finding was entered into the licensees corrective action program as Condition Report 79089. |
| 05000482/FIN-2014002-02 | 2014Q1 | Wolf Creek | Failure to Maintain Licensed Power Limits During Planned Evolutions Affecting Reactivity | A self-revealing non-cited violation, with two examples, of Technical Specification 5.4.1.a, Procedures, was identified for the failure to follow the reactivity management procedures. On two occasions, operators failed to take prudent actions to ensure that reactor power did not exceed the licensed limit of 3565 megawatts thermal while performing activities known to cause power increases. On February 17, 2014, while performing chemical and volume control system inservice check valve testing on the discharge check valve of the train A centrifugal charging pump, operators performed a dilution of the reactor coolant system for normal power maintenance while reactivity was also being affected by the testing of the charging pump check valve, resulting in exceeding 100 percent power. On March 6, 2014, while returning the reactor to full power following data collection on the main turbine control valves, operators used an automatic power ramp to a setpoint of only 3 megawatts below 100 percent, without accounting for the overshoot that would result from the selected ramp rate, resulting in exceeding 100 percent power. In both cases, operators were alerted by an alarm indicating that the 1-minute average power level exceeded 100 percent. The inspectors reviewed station procedure GEN 00-004 Power Operation, and noted a requirement in Attachment A: For pre-planned evolutions that are expected to cause a transient rise in reactor power that could exceed the licensed power level, prudent actions should be taken to reduce power prior to the evolution. Failure to take prudent action to maintain the reactor within licensed power limits prior to performing activities known to cause an increase in reactor power levels is a performance deficiency. The performance deficiency was more than minor because it affected both the configuration control attribute of reactivity control as well as the human performance attribute of procedure adherence of the Barrier Integrity Cornerstone, and impacted the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors screened the finding using the reactivity control screening questions found in Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Section C; question number 3 referred the inspectors to Inspection Manual Chapter 0609, Appendix M, Significance Determination Using Qualitative Criteria. NRC Management performed the qualitative assessment and determined that the finding was of very low safety significance (Green) because the relatively small magnitude of the overpower events, the prompt operator actions to return power to below the licensed limits upon discovery, and the fact that overpower events did not result in any failure of the fuel cladding. The inspectors determined that the finding had a conservative bias cross-cutting aspect in the area of human performance. Specifically, the affected evolutions were known in advance to have positive reactivity impacts; however, operators did not consider reducing power in the case of the check valve testing, nor was a slow approach to the maximum reactor power level used to avoid overshoot during dynamic turbine loading for the turbine valve data collection in order to prevent licensed power levels from being exceeded. |
| 05000317/FIN-2013004-02 | 2013Q3 | Calvert Cliffs | Licensee-Identified Violation | On February 17, 2013, while Unit 2 was in Mode 3 during a refueling outage, CENG personnel identified a pinhole leak at the upper packing leakoff line cap seal weld of pressurizer spray valve 2CV-100F, which constituted RCS pressure boundary leakage. Technical Specifications limiting condition for operation 3.4.13.a, RCS Operational Leakage, limits pressure boundary leakage during plant operation to zero. With any RCS pressure boundary leakage, the technical specifications require the operating unit to be in Mode 3 within 6 hours and to be in Mode 5 within 36 hours. Contrary to the above, based on review of boric acid walkdown data, RCS pressure boundary leakage existed sometime after the last boric acid walkdown conducted in Unit 2 2011 refueling outage and continued during operation for a time longer than allowed by the technical specifications. The inspectors determined that no performance deficiency existed because CENG satisfactorily tested the component using appropriate non-destructive testing prior to installation, identified the boundary leakage through the use of an prescribed monitoring program (boric acid leakage monitoring) and the monitoring frequency was appropriate for the system location (component location inside containment is inaccessible during reactor operation). The inspectors reviewed LER 2013-001-00 and determined that traditional enforcement applies in accordance with IMC 0612, Section 0612-09 and 0612-13 and Enforcement Policy, Section 2.2.4.d, because a violation of NRC requirements existed without an associated significance determination process performance deficiency. This issue was considered to be a Severity Level IV NCV of Technical Specifications limiting condition for operation 3.4.13.a, in accordance with Enforcement Policy, Section 6.1.d. In addition, the inspectors also evaluated this finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings. The inspectors screened the issue and determined that RCS leakage is considered a loss of coolant accident initiator, and evaluated it using the Initiating Event criteria in Appendix A. Assuming worst case degradation, the leakage would not result in exceeding the technical specifications limit for identified RCS leakage (10 gallons per minute) nor would the leakage have likely affected other mitigation systems resulting in a total loss of their safety function. This severity level IV licensee-identified NCV was entered into CENGs CAP as CR-2013-001245. |
| 05000317/FIN-2013004-01 | 2013Q3 | Calvert Cliffs | Inadequate Post-Maintenance Test Associated with an Atmospheric Dump Valve | The inspectors identified an NCV of Technical Specifications 5.4.1, Procedures, for the failure of Constellation Energy Nuclear Group (CENG) personnel to establish, implement, and maintain maintenance requirements associated with No. 21 atmospheric dump valve (ADV). Specifically, CENG personnel failed to perform an adequate postmaintenance test (PMT) in accordance with the work instructions for the No. 21 ADV following maintenance and prior to its return to service. As a result, the valve was returned to service in a condition where its containment isolation function was inoperable. Immediate corrective actions included entering this issue into the corrective action program (CAP). Additional corrective actions taken or planned include training Maintenance shop personnel on writing condition reports (CRs) for all failed PMTs and for Operations to ensure that work orders involving ADVs include post-maintenance operability tests for containment closure. The finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the No. 21 ADV was returned to service in a condition where its containment isolation function was inoperable. In addition, the finding is similar to IMC 0612, Appendix E, Example 5.b, in that, the system was returned to service prior to resolution of the degraded condition. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding does not represent an actual open pathway in the physical integrity of reactor containment. Specifically, there was no loss of steam generator tube integrity. Also, the finding did not involve an actual reduction of hydrogen igniters in the reactor containment. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, CAP component, because CENG staff did not ensure that issues potentially impacting nuclear safety were promptly identified, fully evaluated, and that actions are taken to address safety issues in a timely manner, commensurate with their safety significance. Specifically, CENG staff did not implement a CAP with a low threshold for identifying issues such as writing a CR following the identification that the ADV was degraded. |
| 05000482/FIN-2013003-06 | 2013Q2 | Wolf Creek | Failure to Maintain Complete and Accurate Housekeeping Records | The inspectors identified a Severity Level IV violation of 10 CFR 50.9, Completeness and accuracy of information, for the Wolf Creek Nuclear Generating Stations failure to maintain complete and accurate records required by a license condition. Title 10 CFR 50.9 requires, in part, that information required by statute, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects. Contrary to the above, between October and December 2008, the licensee failed to maintain records required by License Condition 2.C.5 that were complete and accurate in all material respects. Specifically, the Housekeeping Inspection Card for the spent fuel pool area indicated that the inspection had been completed when security access logs indicate that the individual that completed the record had not entered the area. The NRC investigation determined that the assigned individual did not walk down the assigned area, and did not assign a designee to do so (EA-13-084). The failure to maintain records required by License Condition that are complete and accurate in all material respects in accordance with 10 CFR 50.9 was a violation. Because the violation is associated with willfulness and impacted the regulatory process it was evaluated under the traditional enforcement process as set forth in the NRC Enforcement Policy. Since this violation was the result of a willful action, the NRC considers the violation to be more than minor, and therefore, the NRC has classified the violation at Severity Level IV, in accordance with the NRC Enforcement Policy. |
| 05000445/FIN-2012301-01 | 2012Q2 | Comanche Peak | Licensee-Identified Violation | 10 CFR 50.9(a) requires, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material aspects. On June 5, 2012, the NRC gave approval to the licensee to administer a written examination to initial operating license applicants on June 19, 2012. The approval was made based on content of the written examination provided to the NRC on June 4, 2012. In this version of the written examination, Question 71 had been revised based on NRC comments so that it had only one correct answer. Previous draft revisions of the question had two plausible correct answers. The written examination was administered on June 19, 2012. On June 20, 2012, the licensee was conducting the post written examination analysis in accordance with NUREG-1021, ES-403, Section D.3.b. As part of this, the licensee reviews performance on missed questions for training deficiencies and wording problems. While completing this analysis, the licensee identified that the version of Question 71 on the administered written examination was not the version that was approved on June 5, 2012. The licensee notified the NRC of the issue on the same day (June 20, 2012), and completed an extent of condition review that showed that this was the only written examination question provided in the form inconsistent with the questions approved on June 5, 2012. The violation was of very low safety significance because the performance deficiency did not contribute to the NRC making any incorrect regulatory decisions regarding issuance of operating licenses. The licensee entered this issue into their corrective action program as Condition Report CR-2012-006252. |
| 05000482/FIN-2012002-01 | 2012Q1 | Wolf Creek | Inadequate Procedure Causes Lift of Relief Valve and Reactor Coolant Leak During Shutdown | The inspectors reviewed a self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a procedure that failed to restore the reactor coolant pump seal return flow path prior to raising reactor coolant system pressure, which caused the seal return relief valve to lift. During shutdown, reactor coolant pump seal return valve BGHV8100 was shut. On February 12, 2012, Wolf Creek was in Mode 5 with a water-filled (solid) pressurizer at 94 psig. After pressurizer power operated relief valve maintenance, Wolf Creek raised reactor coolant system pressure to 250 psig. With no return path, the relief valve lifted at 150 psig for 15 hours before operators noted an unexplained steady increase in pressurizer relief tank level and re-established the return flow path. Wolf Creek procedures were written to transition straight to refueling, and did not include consideration for maneuvering the plant in Mode 5. This led to shutting valve BGHV8100 without instructions to reopen it before exceeding 150 psig. Wolf Creek subsequently added procedure steps and precautions to reopen the seal return path in Mode 5. The inspectors calculated that approximately 760 gallons of reactor coolant were lost to the relief tank. This issue was placed in the corrective action program as condition report 49021. Failure to align the reactor coolant pump seal return flow path prior to raising reactor coolant system pressure above the relief valve setpoint, creating a leak path, was a performance deficiency. The inspectors determined that this finding impacted the Initiating Events Cornerstone and its objective to limit the likelihood of events that upset plant stability and challenge safety functions during shutdown. Specifically, it impacted the configuration control attribute of shutdown equipment lineup which created an unmonitored intersystem leak. The inspectors used Inspection Manual Chapter 0609, Appendix G, Attachment 1, checklist 4 (cold shutdown, level in the pressurizer, time to boil >2 hours) to evaluate the significance of this finding. A Phase 2 analysis was not needed because the level of inventory was terminated when the normal path was opened and the relief valve reseated. The leak would have terminated itself if the reactor coolant system drained itself to below the pump seal. The finding did not affect reactor coolant system level indication, affect the ability to terminate the leak path, affect the ability to add inventory, or affect the ability to recover residual heat removal if it was lost. Therefore, the finding was determined to be of very low safety significance. The inspectors identified the cause of the finding had a human performance cross-cutting aspect in the area of resources. Specifically, complete and accurate procedures were not provided because Procedure GEN 00-006 did not contain guidance to establish the seal return flow path prior to raising reactor coolant system pressure above 150 psig |
| 05000482/FIN-2012002-04 | 2012Q1 | Wolf Creek | Licensee-Identified Violation | Title 10 of the CFR, Part 50.65(a)(4) requires, in part, that before performing maintenance activities the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Wolf Creek Procedure APF 22B-001, Safety Function Status and Assessment Sumary, revision 3, requires that two or more steam generators have greater than 66 percent wide-range water level to be credited for core decay heat removal risk during Mode 5. Otherwise, the assessment escalates risk due to the absence of one method core decay heat removal. Procedure APF 22B-001 is required to be performed daily. Contrary to the above, on February 6, 2012, the licensee failed to assess risk by identifying entry into a Yellow shutdown risk assessment for core decay heat removal due to three of four steam generators being drained to less than 66 percent wide range level. Specifically, while performing the daily shutdown risk assessment, operators failed to identify that planned work to drain and refill the steam generators for cold wet layup chemistry control was to occur on several steam generators simultaneously. Concurrently, the reactor coolant system power operated relief valves were unavailable for feed and bleed, another method of decay removal. Operators recognized the elevated risk after 8 hours with three steam generators being drained below 66 percent water level. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, operators failed to identify that they had crossed a shutdown risk management threshold, from Green to Yellow. A Region IV senior reactor analyst verified that the finding was of very low safety significance (Green) and the delta-CDF was less than 1E-6. The licensee has entered this issue into their corrective action program as condition report 48775. |