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| Start date | Site | Identified by | Title | Description | |
|---|---|---|---|---|---|
| 05000440/FIN-2018010-01 | 30 September 2018 23:59:59 | Perry | NRC identified | Failure to Correctly Establish Maintenance/Replacement Frequencyfor the WeedTemperature Transmitters In Zone FB-7 | The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations(CFR), Part 50.49(e)(5), Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, for the licensees failure to correctly establish maintenance/replacement frequency for Weed temperature transmitters installed in harsh environment. Specifically, Calculation EQ-115, Qualified Life Calculation for Weed RTD/RTDT and TC Assemblies,incorrectly established a qualified life for Weed temperature transmitters installed in Zone FB-7. The calculation determined that the qualified life for these transmitters in Zone FB-7 as 18.9 years plus accident. However, the calculation failed to account for the accident time and temperature. |
| 05000440/FIN-2018003-01 | 30 September 2018 23:59:59 | Perry | NRC identified | Application of ASME Code Case N5133 for the Emergency Service Water Piping Degradations | The inspectors identified an Unresolved Item concerning the Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, and ASME Code requirements for the ESW piping systems with regards to the licensees application of ASME Code Case N5133, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping, Section XI, Division 1. Updated Safety Analysis Report (USAR) Section 9.2.1 describes that the function of ESW system is to provide a reliable source of water to safety-related components required for normal and emergency reactor operation. USAR Table 3.21, Equipment Classification, delineates that the ESW piping system is safety-related and designed in accordance with the requirements of ASME Section III, Subsection ND (Class 3). The regulation in 10 CFR 50.55a(g) requires, in part, that Class 3 components and their supports meet the requirements of ASME Section XI of the ASME Boiler and Pressure Vessel (BPV) Code or equivalent quality standards. The ASME also publishes Code Cases, which provide alternatives to existing Code requirements. The NRC Regulatory Guide (RG) 1.147 identifies that Code Case N5133 provides acceptable alternatives to applicable parts of Section XI, provided it is used with any identified conditions or limitations. Code Case N5133, Section 2(d) requires that a flaw evaluation shall be performed to determine the conditions for flaw acceptance. Section 3 provides accepted methods for conducting the required analysis. In addition, Section 3 requires, in part, that nonplanar flaws shall be evaluated in accordance with the requirements in 3.2. Additionally, Section 5 requires that an augmented volumetric examination or physical measurement to assess degradation of the affected system shall be performed as follows: (a) From an engineering evaluation, the most susceptible locations shall be identified. A sample size of at least five of the most susceptible and accessible locations, or, if fewer than five, all susceptible and accessible locations shall be examined within 30 days of detecting the flaw. (b) When a flaw is detected, an additional sample of the same size as defined in 5(a) shall be examined. (c) This process shall be repeated within 15 days for each successive sample, until no significant flaw is detected or until 100 percent of susceptible and accessible locations have been examined. On June 13, 2018, a through-wall leakage on the 20 ESW piping was identified in CR 201805504. As a result, the licensee invoked the Code Case to evaluate this flaw and permit the degraded ESW piping system to remain in service for a limited period without repair/replacement. The licensees evaluation involved characterization of this flaw as nonplanar, and subsequently, the methodology as described in Section 3.2 of the Code Case was utilized for this nonplanar flaw. Additionally, the licensee identified the five most susceptible and accessible locations in the ESW system and performed examination in accordance with Section 5(a). From the examination of the five additional locations, another localized wall degradation was identified on the 8 ESW pipe elbow on July 10, 2018. The licensee initiated CR 201806205 to document this condition. The licensee characterized this degradation also as a nonplanar flaw, and this degradation represented approximately 80 percent wall loss from its nominal thickness. During the review of the licensee evaluation of this degraded pipe elbow, the inspectors identified that the methodology as described in Section 3.2 of the Code Case had not been utilized. Instead, the licensee elected to use an alternate methodology to evaluate and disposition for its acceptability. Furthermore, the inspectors identified that the licensee essentially redefined the term flaw in the Code Case to reflect the ASME Section XI, IWA9000 definition of the term defect. The ASME Section XI, IWA9000 defines a flaw as an imperfection or unintentional discontinuity that is detectable by nondestructive examination. It also defines a defect as a flaw (imperfection or unintentional discontinuity) of such size, shape, orientation, location, or properties as to be rejectable. With respect to the Code Case, the licensee essentially restricted the criteria for examination scope expansion only to the flaws that were rejectable; therefore, the licensee had not expanded the scope to perform examination of additional locations in accordance with Section 5(b). In essence, two items are to be further evaluated and addressed: (1) whether the use of methodology not described in the Code Case Section 3.2 was appropriate for evaluation of the nonplanar flaw on the 8 ESW pipe elbow, and (2) whether the stopping of scope expansion for examination as required by the Code Case Section 5(b) was appropriate based on the licensees redefining of the term flaw. In response to the inspectors concern, the licensee initiated CR 201808483, NRC ID: Code Case N5133 Interpretation, September 26, 2018. The licensee also plans to perform examination of five additional locations in November of 2018. This represents an item where the inspectors identified Code interpretation issues that resulted in a disagreement with the licensee. This will require additional review to determine whether a violation exists. Therefore, this issue is considered an unresolved item pending completion of inspector review and evaluation and discussion with the Office of Nuclear Reactor Regulation. Licensee Action: The licensee plans to perform examination of five additional locations in November of 2018. Corrective Action Reference: CR 201808483 |
| 05000440/FIN-2018410-01 | 30 June 2018 23:59:59 | Perry | NRC identified | Security | |
| 05000440/FIN-2018002-01 | 30 June 2018 23:59:59 | Perry | NRC identified | Failure to Control Transient Combustible Materials in a Designated Combustible Control Zone | The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Perry Operating License Condition 2.C(6), Fire Protection, for the licensees failure to control transient combustible materials in a designated combustible control zone within fire area 1AB1g on Auxiliary Building elevation 574 10. Specifically, on May 16, 2018, the inspectors identified transient combustible materials left unattended in the designated combustible control zone in the corridor outside the emergency core cooling system (ECCS) pump rooms, which exceeded the ten pound limit established in the Fire Protection Program document, PAP1910, for ordinary combustibles (loose) in designated combustible control zones without a transient combustible permit. |
| 05000440/FIN-2018001-01 | 31 March 2018 23:59:59 | Perry | NRC identified | Failure to Notify the NRC within 60 Days of a Condition Prohibited by Technical Specifications | The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to report a condition that was prohibited by the plants Technical Specifications to the U.S. Nuclear Regulatory Commission (NRC) within 60 days. Specifically, the licensee did not report a condition that, as determined by the NRC, rendered the Division 2 Diesel Generator (DG) inoperable for a period longer than the Technical Specification allowed completion times of its associated required actions. |
| 05000440/FIN-2017410-01 | 31 December 2017 23:59:59 | Perry | NRC identified | Security | |
| 05000440/FIN-2017008-03 | 31 December 2017 23:59:59 | Perry | NRC identified | Failure to Verify the Capability to Manually Backwash the Emergency Service WaterStrainer during Loss of Offsite Power | The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control,for the failure to verify the capability to manually backwash the emergency service water (ESW) strainer during a LOOP. Specifically, the licensee credited the capability to manually backwash the ESW strainers during a LOOP. However, the associated differential pressure alarm setpoint did not ensure sufficient time to complete this activity because the alarms were set at the same value as the design differential pressure value assumed by the hydraulic calculations. The licensee captured the issue in their CAP as CR-2017-09033, reasonably determined ESW remained operable, and planned to revise the associated calculation and the alarm setpoint to ensure sufficient time to perform the required manual actions during a LOOP.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency did not assure the ESW capability to supply the required minimum flow to its supported components. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the alarm set point was established more than 3 years ago. |
| 05000440/FIN-2017008-02 | 31 December 2017 23:59:59 | Perry | NRC identified | Inadequate Evaluation of Emergency Closed Cooling System Pipe Support | The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to consider all stresses resulting from the emergency closed cooling system as built pipe support 1P42-H1080 connection details. Specifically, the evaluation for the pipe support did not address the impact of rigid connections at both ends of the W8 steel post and of the lateral load on W21 auxiliary steel beam. The licensee captured the issues in their CAP as CR-2017-08986 and CR-2017-09043, reasonably determined the support remained operable, and planned to revise the affected structural analyses.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability,reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.Specifically, the failure to analyze actual pipe configuration and to evaluate the W21 beam did not ensure the emergency closed cooling system and its safety-related supported loads would remain available and capable of providing their accident mitigating function. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. The team determined that this finding had a cross-cutting aspect in the area of human performance because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee did not recognize this latent issue when revising the structural evaluation in 2015. |
| 05000440/FIN-2017008-01 | 31 December 2017 23:59:59 | Perry | NRC identified | Failure to Address the Susceptibility of the Condensate Storage TankLow Level Instrument Lines to Freeze | The team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations(CFR),Part 50, Appendix B, Criterion III, Design Control, and 10 CFR 50.63, Loss of All Alternating Current Power, for the licensees failure to evaluate the capability to transfer the high pressure core spray (HPCS)and the reactor core isolation cooling (RCIC) pumps suction source from the condensate storage tank (CST)to the suppression pool during cold weather conditions. Specifically, (1) monitoring of the CST level instrument lines heat tracing was inadequate to detect a credible common mode failure before the instrument lines would freeze and be rendered inoperable during normal operation, (2)the licensee did not address the condensate (CST) level instrument lines susceptibility to freeze during a cold weather loss of off-site power (LOOP) event with or without a design basis transient or accident, and (3)the licensee incorrectly evaluated the capability to transfer the HPCS pump suction source from the CST to the suppression pool during a cold weather station blackout (SBO) event. The licensee captured the issues within their Corrective Action Program (CAP) as Condition Report(CR) CR-2017-08685, CR-2017-08930, and CR-2017-09006. Corrective actions implemented included: increased the CST level instrument line heat tracing circuit monitoring frequency, revised the affected procedures ensured HPCS and RCIC are adequately aligned to the suppression pool during LOOP design basis events, and ensured a timely transfer of the HPCS and RCIC pump suctions to the suppression pool during a SBO. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability,reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.Specifically, the failure of the HPCS and or RCIC pumps to automatically transfer their suction source from the CST to the suppression pool upon reaching a low CST water level condition could damage the pump(s) thus preventing them to be used to mitigate a transient or accident. A detailed risk evaluation was performed and determined that the finding was of very-low safety-significance (Green). The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the CST instrument lines were designed and the SBO coping strategy during cold weather was established more than 3 years ago. |
| 05000440/FIN-2017003-01 | 30 September 2017 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.5.1, states in part, that the ODCM shall contain the conduct of the Radiological Environmental Monitoring Program (REMP). The ODCM, Revision 20, includes Table 5.1 1 ODCM REMP Locations and Section 3.12.1.c, which states in part, With milk or broadleaf vegetation samples unavailable from one or more of the sample locations...identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. Contrary to the above, as of August 11, 2017, substantive changes to the REMP identified by the licensee in 2015 were not incorporated into the ODCM. Specifically, the licensee identified that a milk sampling location was no longer available and that the expansion of broadleaf vegetation sampling was required. Additionally, the licensee relocated collection sites for water and sediment samples that were not reflected in the ODCM. The licensee documented this issue in CR 2017 08353. The inspectors determined that this REMP issue was of very low safety significance (Green) after reviewing IMC 0609, Appendix D, Public Radiation Safety SDP, dated 23 February 12, 2008. The inspectors determined that this finding was associated with the Environmental Monitoring Program, therefore, the finding screened as Green (verylow safety significance). |
| 05000440/FIN-2017002-02 | 30 June 2017 23:59:59 | Perry | NRC identified | Implementation of Enforcement Guidance Memorandum 11003, Revision 3 | From March 17, 2017, to March 24, 2017, Perry Nuclear Power Plant (PNPP) performed Operations with the Potential to Drain the Reactor Vessel (OPDRV) while in Mode 5 without an operable primary and secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measures to terminate the uncovering of fuel. Secondary containment was required by TS 3.6.4.1 to be operable during OPDRVs. Primary containment was required by TS 3.6.1.10 to be operable during OPDRVS. The required action for these specifications was to suspend OPDRV operations. Therefore, entering the OPDRV without establishing primary and secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B).The NRC issued Enforcement Guidance Memorandum (EGM) 11003, Revision 3, on January 15, 2016, to provide guidance on how to disposition boiling water reactor licensee noncompliance with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to (1) adhere to the NRC plain language meaning of OPDRV activities; (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times; (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a loss of inventory event, (b) maintaining the capability to isolate the potential leakage paths, (c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (d) prohibiting movement of irradiated fuel with the spent fuel storage pool gates removed in Mode 5; and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities.The inspectors reviewed licensee event report (LER) 201700100 for potential performance deficiencies and/or violations of regulatory requirements. The inspectors also reviewed the stations implementation of the EGM during OPDRVs:The inspectors observed that the OPDRV activities were logged in the control room narrative logs, the log entry appropriately recorded the standby source of makeup water designated for the evolutions, and that defense in-depth criteria were in place.The inspectors noted that the reactor vessel water level was maintained at least 22 feet and 9 inches over the top of the reactor pressure vessel flange as required by TS 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designated in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours.The inspectors reviewed Engineering Change documents which calculated the time to drain down during these activities and the feasibility of pre-planned actions the station would take to isolate potential leakage paths during these periods of time. The inspectors verified that the OPDRVs were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the OPDRVs. The inspectors noted that PNPP had in place a contingency plan for isolating the potential leakage path and verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events.The inspectors verified that all other TS requirements were met during the March 17, 2017, to March 24 2017, OPDRVs with primary and secondary containment inoperable.Technical Specification 3.6.4.1 required, in part, that secondary containment shall be operable during OPDRV. Technical Specification 3.6.4.1, Condition C, required the licensee to initiate action to suspend OPDRV immediately when secondary containment is inoperable. Technical specification 3.6.1.10 required, in part, that primary containment shall be operable during OPDRV. Technical specification 3.6.1.10, Condition A, required the licensee initiate action to suspend OPDRV immediately when primary containment is inoperable. From March 17, 2017, to March 24, 2017, PNPP performed OPDRV activities while in Mode 5 without an operable primary or secondary containment. Specifically, the station performed the following OPDRV activities without an operable primary or secondary containment:draining of reactor recirculation loop B; replacement of 18 control rod drive mechanisms (unbolt and install);replacement of six instrument dry tubes;replacement of reactor recirculation pump B seal;replacement of reactor recirculation loop B flow control valve actuator;plugging of drain line appendages on reactor recirculation pump B; andlocal leak rate testing of the reactor water cleanup suction line containment isolation valves.The failure to perform OPDRV activities with operable primary and secondary containments is a violation of TS 3.6.1.10 and TS 3.6.4.1. Because the violation occurred during the discretion period described in EGM 11003, Revision 3, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforcement action for this violation.In accordance with EGM 11003, Revision 3, each licensee that receives discretion must submit a license amendment request within 12 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The inspectors observed thatPNPP is tracking the need to submit a license amendment request as commitment PYL1712101.This LER is closed. This inspection constituted one event follow-up sample as defined in IP 7115305. |
| 05000440/FIN-2017009-01 | 30 June 2017 23:59:59 | Perry | NRC identified | Unsuitable Application of Surge Suppression Diodes in Standby Diesel Generator Control Power Circuitry | Preliminary White . The inspectors identified a finding preliminarily determined to be of low to moderate safety significance (White), and an associated apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) 50, Criterion III , Design Control, for the licensees failure to implement measures for the selection and review for suitability of application of voltage suppression diodes installed in the control circuitry for the Division 2 Standby Diesel Generator, which was a component subject to the requirements of 10 CFR Part 50, Appendix B. Specifically , Engineering Change Package 04 00049 failed to consider the effects of a shorted diode on the control circuitry for the Division 2 Standby Diesel Generator, and instead, introduced new components (diodes) into the control circuitry that resulted in the eventual failure of this safety -related equipment . This rendered the standby diesel generator inoperable and unable to start for longer than its technical specification allowed outage time , which was a violation of Technical Specification 3.8.1, AC Sources -Operating . The licensee documented the issue in CR 2016 13183, and subsequently replaced the failed component and then modified circuitry to remove the replacement diode and the remaining diodes from similar components. The inspectors determined that the licensees failure to evaluate the effects of voltage suppression diode failure on the Standby Diesel Generator control circuit was contrary to the requirements of 10 CFR Part 50, Appendix B , Criterion III and a performance deficiency which was within the licensees ability to foresee and prevent . The inspectors determined that the performance deficiency was of more than minor significance because it was associated with the design control attribute of the mitigating syst ems 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, the design of the Division 2 Standby Diesel Generator control circuit resulted in the inoperability and unavailability of the Division 2 Standby Diesel Generator from April 2, 2015, to November 8, 2016, when the failed diode was replaced. 3 A Significance and Enforcement Revi ew Panel, using IMC 0609, Appendix A, Significance Determination Process for Findings At -Power, dated June 19, 2012, preliminarily determined the finding to be of low -to-moderate safety significance. The inspectors did not identify any cross- cutting aspects associated with this finding because the condition had existed since at least 2007, when the diodes were originally installed in the DC control power circuits, and therefore, was not indicative of current plant performance |
| 05000440/FIN-2017002-01 | 30 June 2017 23:59:59 | Perry | NRC identified | Failure to Notify the NRC within Eight Hours of a Non -Emergency Event that Could Have Prevented the Fulfillment of Multiple Safety Functions | Severity Level IV. The inspectors identified a Severity Level IV Non- Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.72(b)(3)(v)(A) and (D), Immediate Notification Requirements for Operating Nuclear Power Reactors, for the licensees failure to report an event to the NRC within eight hours that at the time of discovery could have prevented the fulfillment of a safety function. Specifically, the licensee did not recognize there was a loss of safety function associated with multiple instrumentation functions as a result of a main steam turbine bypass valve opening at 100 percent reactor power. Therefore, the licensee did not make the required non- emergency eight hour report. After the inspectors questioned the licensees conclusion, the licensee recognized there was indeed a loss of safety function and submitted the eight -hour notification report on May 3, 2017. They also and entered this issue into the corrective action program (CAP) as condition report ( CR) 2017 04939, CR 201704868, and CR 201705022. The failure to make an applicable non- emergency eight -hour event notification report within the required time frame was a performance deficiency. The inspectors determined that traditional enforcement was applicable to the issue because it impacted the NRCs regulatory process. In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d.9 of the NRC Enforcement Policy, this violation was screened as a Severity Level IV violation that was more than minor. In accordance with Inspection Manual Chapter 0612, because this violation involved traditional enforcement and does not have an associated finding that would be considered more- than -minor, a cross-cutting aspect was not assigned to this violation. |
| 05000440/FIN-2017403-02 | 31 March 2017 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | |
| 05000440/FIN-2017001-02 | 31 March 2017 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | In part, 10 CFR 20.1703 (c)(5) states, The licensee shall implement and maintain a respiratory protection program that includes Determination by a physician that the individual user is medically fit to use respiratory protection equipment. Contrary to the above, the licensee identified that an individual wore a powered air purifying respirator (PAPR) three times during the period of March 67, 2017 for the purpose of radiological protection without the required medical determination. This was entered into the licensees corrective action program, CR 201702957, Vessel Technician Wore PAPR Three Times without Being Qualified. The significance of this violation was determined in accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process dated August 19, 2008. This violation was determined to be of very low safety significance (Green), because this violation was not associated with ALARA Planning or Work Controls, there was no overexposure nor substantial potential for overexposure and the ability to access dose was not compromised. |
| 05000440/FIN-2017001-01 | 31 March 2017 23:59:59 | Perry | NRC identified | Failure to Implement Procedures for Combating a Loss of Shutdown Cooling | Green. A finding of very-low safety significance and associated NCV of TS 5.4, Procedures, was identified by the inspectors for the failure to implement procedures for combating a loss of shutdown cooling (SDC). Specifically, the licensee failed to implement its procedure for combating a loss of SDC resulting from emergency service water (ESW) inoperability and during high decay heat load. This finding was entered into the licensees Corrective Action Program to perform analyses for various conditions to identify available alternate methods of decay heat removal and provide associated procedural guidance. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as very-low safety significance (Green) because it was a design deficiency that did not impact the operability or Probabilistic Risk Assessment functionality of any mitigating structures, systems, and components. The inspectors did not identify a cross-cutting aspect associated with this finding because it did not reflect current performance due to the age of the performance deficiency |
| 05000440/FIN-2017403-01 | 31 March 2017 23:59:59 | Perry | NRC identified | Security | |
| 05000440/FIN-2016004-04 | 31 December 2016 23:59:59 | Perry | NRC identified | Failure to Notify the NRC within Eight Hours of a Non-Emergency Event that Could Have Prevented the Fulfillment of a Safety Function | Severity Level IV. The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.72(b)(3)(v)(A) and (D), for the licensees failure to report to the NRC within eight hours, an event or condition that could have prevented the fulfillment of a safety function. The licensees evaluation of this condition, where both trains of the standby liquid control (SLC) system had been inoperable simultaneously, determined that it was not a reportable event. However, the inspectors determined that as described in NUREG 1022, Event Reporting Guidelines 50.72 and 50.73, Revision 3, Section 3.2.7, the licensee had failed to make a non-emergency eight hour report as required by 10 CFR 50.72(b)(3)(v)(A) and (D). The licensee submitted the eight-hour report on December 30, 2016, and entered this issue into the corrective action program (CAP) as CR 201700098. The failure to make an applicable non-emergency eight-hour event notification report within the required time frame was determined to be a performance deficiency. The inspectors determined that traditional enforcement was applicable to this issue because it impacted the NRC's regulatory process. In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d.9 of the NRC Enforcement Policy, this violation was screened as a Severity Level IV violation that was more than minor. In accordance with IMC 0612, because this violation involved traditional enforcement and does not have an underlying technical violation that would be considered more-than-minor, a cross-cutting aspect was not assigned to this violation. |
| 05000440/FIN-2016407-01 | 31 December 2016 23:59:59 | Perry | NRC identified | Security | |
| 05000440/FIN-2016004-02 | 31 December 2016 23:59:59 | Perry | NRC identified | Modifications to Underdrain and Gravity Discharge System Manhole Covers Without a 10 CFR 50.59 Safety Evaluation | Green-Severity Level IV. The inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Test, and Experiments, and an associated finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) with the installation of grated manhole covers, replacing the rubber gasket, watertight manhole covers for the underdrain and gravity discharge systems and did not provide a basis for the determination that this change would not result in a more than a minimal increase in the likelihood of occurrence of a malfunction of a system structure or component important to safety. The licensee entered this issue into the CAP as CR 201611864 and performed a prompt operability determination to show that the underdrain and gravity drain systems remained functional while the engineering change package was developed to support the change and bring the underdrain and gravity discharge systems into compliance with the design basis. The performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone attribute of protection against external factors 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). Per IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, the finding was screened against the Mitigating Systems Screening Questions and determined to be of very low safety significance (Green) because the finding did not cause the underdrain and gravity discharge systems to become inoperable or non-functional. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The inspectors determined it to be of Severity Level IV because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, procedure adherence, in that individuals did not follow processes, procedures, and work instructions. Specifically, a design engineer authorized the permanent modification to be made without the required 50.59 evaluation being completed (H.8). |
| 05000440/FIN-2016004-01 | 31 December 2016 23:59:59 | Perry | Self-revealing | ECC B Heat Exchanger Flow Root Valves Out of Position | Green. A finding of very-low safety significance and associated NCV of TS 5.4.1, Procedures, was self-revealed for the licensees failure to follow valve lineup procedure restoration requirements after an emergency service water (ESW) pump B and valve operability test. Specifically, incorrect valve manipulations of the root valves for 1P42R043B and 1P42R043A flow indicators caused the emergency closed cooling (ECC) heat exchanger B flow to read zero with flow through the heat exchanger. The incorrect flow indication rendered the remote shutdown panel inoperable. The licensee subsequently re-positioned the root valves, 1P42R043B and 1P42R043A, and restored the remote shutdown panel to operable. The licensee entered this issue into the CAP as CR 201612935. The inspectors determined that the performance deficiency for failure to follow procedure was more than minor and thus a finding because it was associated with the Mitigating Systems cornerstone attribute of human performance. The performance deficiency adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has a cross-cutting aspect in the area of human performance, avoid complacency because the licensee failed to ensure that individuals follow processes, procedures, and work instructions. Specifically the individual performing the surveillance did not utilize all the required human performance tools to prevent the error (H.12). |
| 05000440/FIN-2016406-02 | 31 December 2016 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | |
| 05000440/FIN-2016004-03 | 31 December 2016 23:59:59 | Perry | Self-revealing | RCS Pressure Boundary Leakage Operation Prohibited by TSs | Green. A finding of very low safety significance and an associated non-cited violation (NCV) of Technical Specification (TS) 3.4.5, RCS Operational Leakage, was self-revealed when the licensee operated with reactor coolant system (RCS) pressure boundary leakage, as a result of the failure of the weld connecting the root appendage of the vent line on the recirculation loop A discharge valve, between January 19, 2016, and January 24, 2016, which is a condition prohibited by TS. The licensee entered this issue into the Corrective Action Program (CAP) as Condition Report (CR) 201601071 and performed a significant condition adverse to quality root cause evaluation due to a principal safety barrier being seriously degraded, replaced the vent line appendage on the recirculation loop A discharge valve with a more robust pipe and cap, and developed plans to replace ten additional vent and drain line appendages on the reactor recirculation loops prior to the end of the 1R17 refueling outage in 2019. The inspectors determined that the licensees operation with RCS pressure boundary leakage, a condition prohibited by TSs, was a performance deficiency requiring evaluation. The inspectors determined that the finding was more than minor because it adversely impacted the Initiating Events cornerstone attribute of equipment performance-barrier integrity, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined this finding was of very low safety significance because the leak would not have exceeded the RCS leak rate for a small loss-of-coolant accident (LOCA) and would not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function. The inspectors concluded that this finding had no additional cross-cutting aspects than what was discussed in Inspection Report 0500440/2016001. |
| 05000440/FIN-2016406-01 | 31 December 2016 23:59:59 | Perry | NRC identified | Security | |
| 05000440/FIN-2016009-02 | 30 September 2016 23:59:59 | Perry | NRC identified | Failure to Establish a Periodic Maintenance Program for Communications Equipment Associated with FLEX | A finding of very low safety significance was identified by the inspectors for failing to establish period tasks to check the operation of recently installed FLEX related communications equipment in accordance with the Perry Nuclear Power Plant FLEX Final Integrated Plan Report. The licensee entered this issue into the corrective action program as CR201609746 and 201609747 to develop the appropriate periodic maintenance tasks. The finding was determined to be more than minor because it was associated with the Emergency Preparedness Cornerstone Attribute of Facilities and Equipment which includes Maintenance Surveillance and Testing of Facilities, Equipment and Communications Systems. Specifically, communications equipment, particularly batteries, degrade over time and without periodic checks to verify functionality, the equipment might not be available for response to a potential accident. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, appendix M, Significance Determination Process using Qualitative Criteria, informed by draft appendix O, Significance Determination Process for Mitigating Strategies and Spent Fuel Pool Instrumentation (Orders EA-12-049 and EA-12-051). The finding screened as very low safety significance, Green, because the inspectors answered no to all Appendix O questions. This finding has a cross-cutting aspect in the area of Human Performance, Work Management because a task to create the activities was initiated, but the completion date was postponed well past the date at which the licensee declared compliance with mitigating systems orders. |
| 05000440/FIN-2016009-01 | 30 September 2016 23:59:59 | Perry | NRC identified | Failure to Implement a Periodic Replacement Program for FLEX Hoses | A finding of very low safety significance was identified by the inspectors for failing to establish a periodic replacement program for the high-temperature rated hoses used during a mitigating strategy for suppression pool cooling. Specifically, the licensee failed to create a periodic replacement program for high temperature FLEX hoses based on the vendor recommendation of a six year shelf-life or justify deviation from the recommendation. The licensee entered this issue into the corrective action program as CR201609776 with an action to generate the appropriate repetitive task for periodic replacement of the high-temperature rated hose. No violation of NRC requirements were identified. This performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects 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), and is therefore a finding. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, appendix M, Significance Determination Process using Qualitative Criteria, informed by draft appendix O, Significance Determination Process for Mitigating Strategies and Spent Fuel Pool Instrumentation (Orders EA-12-049 and EA-12-051). The finding screened as very low safety significance, Green, because the inspectors answered no to all Appendix O questions. This finding had a cross-cutting aspect of Procedure Adherence in the area of Human Performance because the licensee failed to follow procedural guidance to replace hoses based on vendor recommendations. |
| 05000440/FIN-2016007-01 | 30 June 2016 23:59:59 | Perry | NRC identified | Failure to Document 50.59 Evaluation for Replacement of a Manual Action with an Automatic Action | The inspectors identified a Severity Level IV, NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50.59, Changes, Tests, and Experiments, having very-low safety significance (Green) for failure to document the basis for performing a plant modification where a manual operator action was replaced with an automatic action. Specifically, the licensee did not evaluate whether adding a safety-related function to a nonsafety-related component was within the licensing basis of the facility. The inspectors determined that the failure to perform a 10 CFR 50.59 evaluation for Plant Modification 11-0794 was contrary to 10 CFR 50.59(d)(1) and was a performance deficiency. The performance deficiency was determined to be more-than-minor and a finding, because the finding impacted mitigating systems cornerstone attribute of Design Control 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, this plan modification added a Safety-Related function to a Nonsafety-Related component and, therefore, impacted the availability, reliability, and capability of the Safety-Related Battery Room ventilation system and the Safety-Related Motor Control Center, Switchgear, and Miscellaneous Electrical Equipment Area ventilation system. In addition, the associated violation was determined to be more-than-minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined that finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process. Using Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined that the finding affected the Mitigating Systems cornerstone. As a result, the inspectors evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, for the Mitigating Systems cornerstone. The inspectors answered No to question A.4 in Exhibit 2 Mitigating System Screening Questions. Specifically, the inspectors determined the finding did not represent an actual loss of the Battery Room ventilation system or Motor Control Center, Switchgear, and Miscellaneous Electrical Equipment Area ventilation system because the automatic action had not been implemented at the time of the finding. Therefore, the inspectors determined the significance of this finding to be of very-low safety significance (Green). In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding). The inspectors determined the finding was associated with the cross-cutting aspect of Procedure Adherence in the area of Human Performance, because the licensee failed to follow the screening criteria in Attachment 2 of Procedure NOBP-LP-4003A, FENOC 10 CFR 50.59 User Guidelines. |
| 05000440/FIN-2016002-03 | 30 June 2016 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.5.1, Offsite Dose Calculation Manual (ODCM), requires, in part, that radioactive effluents control information be contained within the ODCM. ODCM, Revision 20, requires that the liquid radioactive waste to essential service water radiation monitor have periodic channel functional tests performed. Contrary to the above, on December 13, 2015, it was identified by the licensee that this test, which was required on September 13, 2015, had not been performed. On August 26, 2015, the licensee identified that the monitor would not pass the required functional test. The licensee incorrectly deferred the required channel functional test. This was identified by the operation department staff. The licensee documented this issue in CR201516711, December 13, 2015. The finding was determined to be of very-low safety significance (Green) because it was not a failure to implement the Effluent Program, nor did public dose exceed Appendix I or 10 CFR 20.1301(e) criteria. |
| 05000440/FIN-2016002-02 | 30 June 2016 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | Title 10 of the CFR, Part 50.65(a)(4) states, in part, Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, the licensee identified that it failed to perform the required risk assessment prior to commencing maintenance activities. Specifically, on April 28, 2016, the licensee racked out the L1006 breaker, which was the alternate supply to bus L11 from the Unit 1 startup transformer. The At-The-Controls licensed reactor operator then questioned whether or not an availability log entry was required for this breaker being racked to the disconnect position. At that time, the unit supervisor ran the probabilistic risk assessment (PRA) program for the L1006 breaker being racked to the disconnect position with the plant in its current configuration. The result of the PRA resulted in an increase in risk from Green Risk to Yellow Risk. The Unit Supervisor stopped the evolution and directed the breaker to be racked back in. After an evaluation of the schedule was performed and permission was received from the general plant manager to continue with the plant in Yellow PRA Risk, breaker L1006 was racked out and preventive maintenance proceeded. The finding was determined to be of very low safety significance because the risk deficit incremental core damage probability risk assessment increase of 6.6E10 was less than the 1E6 threshold. The licensee initiated CR201606093 to address this issue. |
| 05000440/FIN-2016007-02 | 30 June 2016 23:59:59 | Perry | NRC identified | Use of Unapproved Standard for Site Flooding Modifications and Analysis | The inspectors identified a Severity Level IV, NCV of 10 CFR 50.59, Changes, Tests, and Experiments, having very-low safety significance (Green) for the licensees failure to conclude that site flooding modifications and associated analysis included a standard that resulted in a departure from the method of evaluation as described in the Updated Final Safety Analysis Report. Specifically, the licensee used a new method for evaluation of design basis flooding at Perry Nuclear Power Plant that is different from the method described in the Updated Final Safety Analysis Report and not approved by the NRC. The inspectors determined that the licensees use of an unapproved methodology for site flooding modifications and associated analysis that constituted a departure from a method of evaluation was contrary to 10 CFR 50.59(c)(2)(8) and was a performance deficiency. Specifically, the licensee used a new method for evaluation of design basis flooding at Perry Nuclear Power Plant that is different from the method described in the Updated Final Safety Analysis Report and not approved by the NRC. The performance deficiency was determined to be more-than-minor, and a finding, because it affected the cornerstone attribute of protection against external factors and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In addition, the associated violation was determined to be more-than-minor because the inspectors determined that there was a reasonable likelihood that the changes would have required prior NRC approval. The inspectors determined that finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process. Using Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined that the finding affected the Mitigating Systems cornerstone. As a result, the inspectors evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, for the Mitigating Systems cornerstone. The inspectors answered Yes to question A.1 in Exhibit 2 Mitigating Systems Screening Questions. Specifically, the inspectors determined the finding did not result in systems, structures, and components not being able to maintain their operability or functionality. Therefore, the inspectors determined the significance of this finding to be of very-low safety significance (Green). In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding). The inspectors determined that this finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Problem Identification, for the licensees failure to identify issues completely, accurately, and in a timely manner. Specifically, the licensees 50.59 review committee failed to accurately identify the methodology change concern in Evaluation 14-01234 during a review documented in CR2015-14025. |
| 05000440/FIN-2016002-01 | 30 June 2016 23:59:59 | Perry | NRC identified | Failure to Comply with ODCM During Liquid Effluent Discharge | A finding of very low safety significance, and an associated NCV of Technical Specification (TS) 5.5.1 was identified by the NRC inspectors for the failure to follow Offsite Dose Calculation Manual (ODCM) requirements during the execution of a liquid effluent discharge. The license entered this event into their CAP as CR201607572 and the individual was coached regarding procedure compliance. The inspectors determined that the performance deficiency was more than minor because the issue impacted the program and process attribute of the Public Radiation Safety cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, on February 1, 2016, a liquid effluent discharge was performed with the radwaste to essential service water discharge monitor inoperable and without the required independent verification of release rate calculations. The finding was determined to be of very low safety significance (Green) because it was not a failure to implement the Effluent Program, nor did public dose exceed Appendix I or Title 10 of the Code of Federal Regulations (CFR), Part 20.1301(e) criteria. The inspectors concluded that the finding had a cross-cutting aspect in the human performance area of procedure adherence because procedures for this task were not followed (IMC 0310, H.8). |
| 05000440/FIN-2016007-03 | 30 June 2016 23:59:59 | Perry | NRC identified | Failure to Comply With ASME Code Requirements for Repair on Code Class 1 Component | A finding of very-low safety significance (Green) and associated NCV of 10 CFR 50.55a(g)(4) was identified by the inspectors for the licensees failure to maintain the American Society of Mechanical Engineers (ASME) Code Class 1 component in accordance with ASME Code Section XI requirements. Specifically, the licensee failed to measure and document the method of measuring the cavity created after removal of indications on the reactor water clean-up line prior to return to service. The inspectors determined that the licensees failure to maintain the ASME Code Class 1 component in accordance with ASME Code Section XI requirements was a performance deficiency. This performance deficiency was found to be more-than-minor, and a finding, because the performance deficiency, if left uncorrected could become a more significant safety concern. Specifically, absent NRC identification, the licensee would not have questioned the potential challenge to component functionality since the cavity measurements were not performed. This condition could potentially lead to the failure of the reactor water clean-up bottom head drain, which in turn, could lead to a potential loss of reactor coolant. The inspectors reviewed the finding using Attachment 0609.04, Initial Characterization of Findings, Table 3 SDP Appendix Router. The inspectors answered No to the question in Section A of Table 3 and therefore the finding was evaluated using the SDP in accordance with IMC 0609, The Significance Determination Process (SDP) for At-Power Operations, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors answered No to the questions in Exhibit 1 and determined this finding to have a very-low safety significance (Green). The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Design Margin, for the licensees failure to maintain equipment within design margins. Specifically, the licensee staff failed to ensure that metal removal performed on an ASME Code Class 1 component did not result in a condition where the minimum design wall thickness of the component was compromised, and therefore, failed to ensure design margin was maintained. |
| 05000440/FIN-2016001-02 | 31 March 2016 23:59:59 | Perry | Self-revealing | Failure to Control Welding and Inspection Activities to Maintain Reactor Coolant System Integrity | A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, was self-revealed on January 24, 2016, for the licensees failure to control welding and inspection activities during the replacement of the reactor recirculation loop A pump discharge valve vent line during the 2015 refueling outage. When identified as the source of reactor boundary leakage in January 2016, the licensee determined that the weld did not meet the requirements on the design drawing and that the quality control (QC) inspection should have identified the non-conforming weld. The issue was entered into the licensees corrective action program as CR 201601071. Corrective actions included installation of an alternative pipe and cap to replace the failed vent line appendage, plugging and capping of the reactor recirculation loop A flow control valve vent line appendage and performed a weld build up on the reactor recirculation loop B flow control valve vent appendage line. The inspectors determined that the licensees failure to control welding and inspection activities was a performance deficiency that was determined to be more than minor and thus a finding, because it was associated with the Initiating Events cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance because it was determined that after a reasonable assessment of degradation, the leak would not have exceeded the reactor coolant system leak rate for a small-break loss of coolant accident (LOCA) and the leak would not have affected other systems used to mitigate a LOCA (e.g., an interfacing system LOCA). This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, the licensee failed to provide additional precautions, controls, and oversight for the personnel performing the welding activities, inspection activities, and supervisory activities, such that the welder, QC inspector, and supervisor were able to complete a weld that met the requirements of the design drawing and to perform an adequate inspection of the weld to determine that it met the acceptance criteria established by the design drawing (IMC 0310, H.1). |
| 05000440/FIN-2016008-03 | 31 March 2016 23:59:59 | Perry | Self-revealing | Hardcard Development Failed to Follow Procedure Change Process | A self-revealed finding and an associated NCV of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to prescribe instructions appropriate to the circumstance into procedures for an activity affecting quality. Specifically, the licensee failed to incorporate instructions into procedures to fill and vent all portions of the reactor water level reference leg purge system. This issue has been entered the issue into the CAP as CR 201602716 to provide a process for the activities. The failure to prescribe instructions appropriate to the circumstance into procedures for an activity affecting quality was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenged critical safety functions during shutdown as well as power operations and was therefore a finding. Specifically, gas left in the reactor water level instrument reference leg purge system during maintenance equipment alignment was known to have the potential to interfere with the proper operation of pressure and level indicators relied upon for safety functions, as documented in Generic Letter 9303. The finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident (LOCA), cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined this finding had a cross-cutting aspect of challenge the unknown in the human performance area where individuals stop when faced with uncertain conditions and risks are evaluated and managed before proceeding. Specifically, the technicians involved in the April 18, 2015, system recovery activities did not stop when faced with an uncertain condition, communicate with supervisors, nor consult system experts to resolve the condition prior to continuing work activities. Since this condition was not placed into the corrective action process at the time, no further consideration was given to venting the reference leg portion of the reactor water level reference leg purge system (IMC 0310, H.11). |
| 05000440/FIN-2016001-01 | 31 March 2016 23:59:59 | Perry | Self-revealing | Failure to Properly Implement System Operating Instructions to Maintain Control of Reactor Pressure Vessel Level | A finding of very low safety significance and an associated non-cited violation (NCV) of Technical Specification (TS) 5.4.1., Procedures, was self-revealed on January 24, 2016, when an unplanned automatic reactor protection system (RPS) actuation occurred as a result of the licensees failure to correctly implement the steps outlined in procedure SOIC34, Feedwater Control System, Section 4.2.12.c to balance inservice flow controller outputs. Specifically, while in the process of reducing power to allow for a drywell entry to determine the location of an unidentified leak into the drywell floor drain sump, the operators failed to control reactor pressure vessel water level during shifting of feedwater pumps from a turbine-driven reactor feed pump to the motor-driven reactor feed pump, resulting in a RPS actuation initiated on reactor vessel water Level 8, shutting down the reactor. Following the reactor scram, the licensee took immediate actions to restore and maintain RPV water level in accordance with procedure ONIC711, Reactor Scram, Revision 20. The issue was entered into the licensees corrective action program as CR 201601063. The licensees failure to properly implement the steps in the procedure was a performance deficiency that was determined to be more than minor and thus a finding, because it was associated with the Initiating Events cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance because it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the licensee failed to provide adequate, procedural guidance on when to conduct the feedwater pump shift (IMC 0310, H.1). |
| 05000440/FIN-2016008-04 | 31 March 2016 23:59:59 | Perry | NRC identified | Failure to Maintain Traceability of Safety Related Fuses | The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 CFR 50, Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components, for the licensees failure to assure that identification of items was maintained by appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item. Specifically, the licensee failed to maintain traceability of safety related fuses installed in safety related systems. The licensee has entered this issue into the CAP as CR 201602048 and CR 201602258. Corrective actions being performed by the licensee include evaluating implementation of procedure NOPWM4300 for documenting use of parts in safety related systems and issuing work orders to determine where the potentially defective fuses were installed in the Division 2 and 3 safety related buses for replacement. The inspectors determined that the failure to assure that identification of items was maintained by appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item was a performance deficiency. Specifically, the licensee failed to maintain traceability of safety related fuses installed in safety related systems. The performance deficiency was more than minor because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, identification and control measures are designed to prevent the use of incorrect or defective materials, parts or components which could render safety systems inoperable. Additionally, the performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and was, therefore, a finding. The finding was determined to be of very low safety significance because the finding was not a deficiency affecting the design or qualification of a mitigating structure system or component, did not represent a loss of system safety function, did not represent an actual loss of function of a single train or two separate trains for greater than its allowed outage time, and did not represent an actual loss of safety function of one or more non-technical specifications trains of equipment during shutdown for equipment designated as high safety significant for greater than 24 hours. The inspectors determined this finding had a cross-cutting aspect of documentation in the human performance area where the organization creates and maintains complete, accurate and up-to-date documentation. Specifically, a review by the licensee of existing work orders that may have utilized the fuses did not clearly document if the fuses were installed, returned to the warehouse or scrapped (IMC 0310, H.7). |
| 05000440/FIN-2016403-01 | 31 March 2016 23:59:59 | Perry | NRC identified | Security | |
| 05000440/FIN-2016001-03 | 31 March 2016 23:59:59 | Perry | Self-revealing | Failure to Take Actions to Prevent a Loss of Safety Function during Reactor Recirculation Pump Downshift | A finding of very low safety significance and an associated NCV of TS 5.4.1, Procedures, was self-revealed on January 24, 2016, when a loss of safety system function occurred as a result of the operators failing to take steps to prevent all operable average power range monitors (APRMs) from becoming out of specification in the non-conservative direction after a recirculation pump shift to slow speed. Specifically, while in the process of reducing power to allow for a drywell entry at low power, the recirculation pumps were shifted and all operable APRMs went out of specification low, which is the non-conservative direction. The operators immediately declared the APRMs inoperable and took actions to restore the operability of at least one APRM in each channel. The issue was entered into the licensees CAP as CR 201601058. The licensees failure to take action to prevent all operable APRMs from going out of calibration low, despite understanding the cause, was determined to be more than minor and thus a finding, because it was associated with the Mitigating Systems cornerstone attribute of human performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it did not result in the loss of reactivity control systems beyond a single trip signal function and did not result in a mismanagement of reactivity by the operators. This finding has a cross-cutting aspect in the area of human performance, avoid complacency, for knowing that the APRMs would go out of calibration because of the pump shift but without regard for the inherent risk while expecting the successful outcome that at least one would stay in calibration without any consideration of potential actions that could have been taken to prevent the loss of safety function and reportable condition (IMC 0310, H.12). |
| 05000440/FIN-2016008-02 | 31 March 2016 23:59:59 | Perry | NRC identified | Failure to Provide Instructions to Completely Vent Reference Legs | The inspectors identified a finding of very low safety significance and an associated non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to follow fleet procedure NOPSS3001, Procedure Review and Approval, and to ensure that a newly developed hardcard was properly reviewed and approved prior to implementation. Specifically, the licensee characterized the hardcard development and implementation as only an administrative change, and was thereby exempted from the fleet procedure review process for new procedures. The licensee entered this finding into the corrective action program (CAP) as condition report (CR) 201603033 and planned to perform a causal review to ensure that actions taken in response to information provided in operations administrative instruction, OAI1703, Hardcards, have received appropriate review under 10 CFR 50.59. The inspectors determined that the failure to follow the licensees fleet and site-specific procedures to ensure that a newly developed hardcard was properly reviewed and approved prior to implementation was a performance deficiency. The performance deficiency was more than minor because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing review and approval activities in accordance with established procedures, the licensee might unintentionally challenge the operators by requiring equipment manipulation that impose unnecessary plant transients, which would result in unwarranted challenges to safety related equipment. Additionally, the performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, and was therefore a finding. The finding was determined to be of very low safety significance because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding had a cross-cutting aspect of conservative bias in the human performance area where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, when the licensee determined to develop the hardcard procedure as an administrative change, the decision precluded the opportunity for the licensee to properly evaluate that the procedure actions did not adversely impact existing station procedures and equipment (IMC 0310, H.14). |
| 05000440/FIN-2016008-01 | 31 March 2016 23:59:59 | Perry | NRC identified | Coincidence Logic to Preclude Spurious Trips of the Offsite Power Source | The inspectors identified an unresolved item (URI) concerning the installed designed of the safety-related 4-kilovolt under voltage protection scheme. On February 11, 2016, while the plant was shut down, a fuse failure in a bus voltage detection scheme resulted in the actuation of associated under voltage relays and a trip of the safety related EH11 bus. With the under voltage condition locked in, the Division 1 safety-related equipment remained unavailable when the Division 1 EDG started and powered up the EH11 bus. As a consequence of the under voltage scheme, the ESW pump for the EDG was not available to provide cooling water and the EDG was manually shut down by operators to prevent damage to the diesel engine. By letter dated June 3, 1977, Statement of Staff Positions Relative to Emergency Power Systems for Operating Reactors (Agencywide Document Access and Management System (ADAMS) Accession No. 8111230342), the NRC requested all licensees, including Perry Nuclear Power Station, to assess the susceptibility of Class 1E electrical equipment to sustained degraded voltage conditions from offsite power sources and to the interaction between the offsite and onsite emergency power systems. In this same letter, the NRC requested that licensees compare the current design of the emergency power systems at plant facilities with the NRC staff positions and that licensees propose plant modifications, as necessary, to meet the NRC staff positions, or provide a detailed analysis which shows that the facility design has equivalent capabilities and protective features. The NRC staff subsequently issued BTP PSB1, Adequacy of Station Electric Distribution System Voltages, Revision 0, dated July 1981 (Appendix A to Standard Review Plan Chapter 8), which provided additional guidance on technical requirements of the degraded voltage scheme. The NRC staff in addressing Perrys design for conformance with BTP PSB1, stated in Section 8.2.4.1 of the Safety Evaluation Report (SER) for Perry (ADAMS Accession No. 8211120305 dated May 31, 1982) that the applicant states in the FSAR that degraded voltage conditions for the Class 1E power system are detected by an under voltage protection system on each division... The system includes specific coincident logic for each bus. The staff also stated that the applicant has committed to provide the final design of the first and second level under voltage protection of the safety equipment in conformance with the staff position, before plant startup. Therefore, the staff finds this to be acceptable pending confirmation of the set point values and analysis. The staff identified this item as Confirmatory Issue (34) in the Section 1.0 of the SER. In Section 8.2.5 of this SER the staff also stated that on the basis of this review..., the staff concludes that the offsite power system for Perry meets the requirements of GDC 17 and 18 and is, therefore, acceptable. By letters dated June 8, 1982, and August 26, 1982, (ADAMS Accession No. 8206170298 and 820310416) the licensee provided details of the degraded voltage relay scheme to be utilized at Perry. In the August 26, 1982, letter, the licensee provided details associated with set point for the first level of under voltage protection. Specifically the licensee stated that the first level of voltage protection had been changed to trip at 75 percent of motor rated voltage instead of 86 percent and the three second fixed time delay would still remain in effect. The licensee also stated the diesel generator start signal after the 15 second time delay for the second level of under voltage had been eliminated. The licensee attached sketches to clarify the logic for a typical bus under voltage protection scheme. However, the licensee did not provide a detailed description of how the attached sketch satisfied the requirements of BTP PSB1. The August 1982 letter again states that the final design and set points will be established after a review of onsite preoperational test results. In a Supplemental Safety Evaluation Report (SSER 2) (ADAMS Accession No. 8301280169 dated January 13, 1983), the NRC staff acknowledged that based on information provided in the August 26, 1982 letter, the relays in the loss of power protection relays were arranged on a two-out-of-two coincident logic to initiate a timer with a 3-sec time delay. The SSER also acknowledged that in the degraded grid voltage protection, the relays were arranged in a two-out-of-two coincident logic to initiate two separate time-delay relays. The staff stated that the applicants final design of the first and second level under voltage protection of safety equipment was acceptable; thus Confirmatory Issue (34) listed in Section 1.10 of the SER, was satisfactorily resolved. The licensee revised the USAR to document conformance with BTP PSB1 and the requirements of Institute of Electrical and Electronics Engineers (IEEE) Std. 2791971, Criteria for Protection Systems for Nuclear Power Generating Stations. Specifically, USAR Section 8.3.1.1.2.9.2 states that under voltage protection shall include coincidence logic on a per bus basis to preclude spurious trips of the offsite power source. This is in conformance with staff guidance stating that improper voltage protection logic can itself cause adverse effects on the Class 1E systems and equipment such as spurious load shedding of Class 1E loads from the standby diesel generators and spurious separation of Class 1E systems from offsite power. The inspection team noted that the licensees installed design utilizes a single voltage sensor that is shared between two relays in the under voltage trip logic. As a consequence of a shared voltage sensor between both under voltage relays, the overall under voltage protection did not constitute a coincidence logic as stated in BTP PSB1. A single malfunction in the voltage sensing circuit resulted in a trip of the offsite power source and precluded the onsite power source from performing its safety related functions. As evidence by the February 11, 2016, event, a secondary potential transformer fuse failure caused the EH11 bus to separate from offsite power (even though there were no deficiencies in offsite power voltage) and resulted in shutdown of the division 1 EDG bus due to ESW being unavailable. The inspectors believe that this design is deficient in that there is no coincident logic to ensure that spurious trips are precluded as delineated in BTP PSB1 Section B.1.c. The licensee agrees that the issue relative to the failure of a single fuse resulting in actuation of both relays, satisfying the under voltage protection scheme logic, represented a design vulnerability. However, the licensee contended that this original design had been maintained, and was approved by the NRC during the initial licensing. Specifically, the licensee concluded, in a white paper addressing the SIT concerns, that The current design of the potential transformers fusing and the under voltage relays for the division 1 4160 V ESF bus is reflective of the original design and licensing basis. The NRC Staff reviewers clearly used PSB1 in their review of the PNPP licensing and design bases. They were aided by the schematics contained in the following letter (First Energy Nuclear Operating Company (FENOC) letter docket Nos. (50440; 50441), dated August 26, 1982). They recognized and approved of the relays arrangement in a two-out-of-two coincident logic. (This is identified on the second schematic - reference Sheet 2 of 2 on lines 19 and 21, which address the time delay relays.) This design is reflective of PNPP's current design and is consistent with the original PNPP licensing basis. Additionally, the NRC stated in the SER that Perry meets the requirements of GDC 17 and 18. This issue is considered an unresolved item (URI 05000440/20160801) pending receipt of clarification of the design basis with respect to 4 Kilovolt safety bus under voltage protection scheme. |
| 05000440/FIN-2015008-03 | 31 December 2015 23:59:59 | Perry | NRC identified | Failure to Provide Adequate Guidance to Override Spurious CO2 Initiation Signal in the Diesel Generator Rooms | The inspectors identified a finding of very low safety significance (Green), and an associated NCV of TS Section 5.4.1.a for the licensees failure to have adequate procedural guidance in their fire response procedure. Specifically, Procedure ONI-P54, Fire, Revision 19 did not list all the fire areas where a potential fire induced spurious carbon dioxide (CO2) initiation in the emergency diesel generator (EDG) room could occur. The licensee entered this issue into their CAP, and established hourly fire watches for the affected areas. The inspectors determined that the performance deficiency was more than minor because a fire in any of the affected fire zones could damage circuits for the nonsafety-related CO2 systems for the EDG rooms causing a potential spurious CO2 initiation in the diesel rooms and affecting the operation of the ventilation fans and dampers in the diesel rooms. The finding was of very low safety significance because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. The finding did not have a cross-cutting aspect associated with it because it was not reflective of current performance. |
| 05000440/FIN-2015403-06 | 31 December 2015 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | |
| 05000440/FIN-2015004-02 | 31 December 2015 23:59:59 | Perry | NRC identified | Liquid Effluent Calibration | The inspectors identified that the efficiency calibration for the liquid effluent radiation monitor, 0D17K0606, could not be located. The licensee performed a new efficiency determination on the monitor during digital modification upgrade on 2006 for all effluent monitors. According to the licensee, the calculated count rate using a new standard National Institute of Standards and Technology traceable sources indicated a close approximation to the liquid detector count rates data determined during the detector initial (primary) calibration. To date, the licensee was unable to provide the initial calibration paperwork indicating that the calibration count rates for the detector efficiency determinations were correlatable. The inspectors attempted to assess whether the original standard count rates for efficiency determination were correlatable to the initial calibration paperwork; however, this assessment could not be completed within this inspection period. The issue remains under review by the U.S. Nuclear Regulatory Commission (NRC) pending further information from the licensee, and is categorized as an Unresolved Item (URI) pending completion of that NRC review. |
| 05000440/FIN-2015008-01 | 31 December 2015 23:59:59 | Perry | NRC identified | Failure to Ensure that Systems, Structures, and Components Necessary to Achieve and Maintain Hot Shutdown Conditions were Free of Fire Damage without Repair Actions | The inspectors identified a finding of very low safety significance (Green), and associated NCV of license condition 2.C(6) for the licensees failure to ensure that systems, structures, and components necessary to achieve and maintain hot shutdown conditions were free of fire damage. Specifically, the licensee did not ensure that circuits associated with the emergency closed cooling (ECC) heat exchanger A temperature control valve 1P42-F665A were free of fire damage for a fire in the control room and instead relied on lifting leads and replacing fuses to take manual control of the valve. The licensee entered the issue into their CAP, and credited the existing repair activities in the procedure. The inspectors determined that the performance deficiency was more than minor because a fire in the control room could result in the licensee losing the ability to remotely control the ECC heat exchanger A temperature control valve and needing to take manual control of the valve. The finding was of very low safety significance because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. The finding did not have a cross-cutting aspect associated with it because it was not reflective of current performance. |
| 05000440/FIN-2015004-01 | 31 December 2015 23:59:59 | Perry | NRC identified | Failure to Ensure Required 3 Hour Fire Barriers (gypsum board walls) Were In-Place | The inspectors identified a finding of very low safety significance and an associated NCV of Perry Operating License Condition 2.C(6), Fire Protection, for the licensees failure to maintain a three-hour fire barriers as required by the Updated Safety Analysis Report (USAR). Specifically, the inspectors identified a through-wall hole, approximately two feet wide and two feet tall in the common wall between the Unit 2, Division 1 and Division 2, direct current (DC) switchgear rooms and another hole, approximately one foot wide and one foot tall between the Unit 2, Division 2 DC switchgear room and the outside hallway. The two through-wall holes were determined to be a performance deficiency associated with compliance to the licensees fire protection program because the walls are described in the USAR as three-hour fire barriers for the rooms in question. The performance deficiency was more than minor; and thus a finding, because it was associated with the Protection Against External Factors 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. The inspectors determined that the finding was of very low safety significance through analysis of the issue as a fire confinement problem and the fact that the reactor would still be able to reach and maintain safe shutdown despite the deficiency. The inspectors identified no cross-cutting issues associated with this finding because the condition has existed since at least July 2011, and therefore, is not indicative of current plant performance. |
| 05000440/FIN-2015403-05 | 31 December 2015 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | |
| 05000440/FIN-2015408-01 | 31 December 2015 23:59:59 | Perry | Licensee-identified | Licensee-Identified Violation | |
| 05000440/FIN-2015403-01 | 31 December 2015 23:59:59 | Perry | NRC identified | Security | |
| 05000440/FIN-2015403-03 | 31 December 2015 23:59:59 | Perry | NRC identified | Security | |
| 05000440/FIN-2015008-02 | 31 December 2015 23:59:59 | Perry | NRC identified | Failure to Inspect Penetration Seals Within the Required Time Frequency | The inspectors identified a finding of very low safety significance (Green), and an associated NCV of license condition 2.C(6) for the licensees failure to adequately implement and maintain surveillance procedures and work processes associated with fire barrier and penetration seal inspections. Specifically, the licensee failed to perform fire barrier penetration seal inspections for 42 penetration seals at least once per 15 years (plus an additional 25 percent grace period) as required by the Fire Protection Program. The licensee entered the issue into their CAP, and will inspect the accessible portions of the barriers and will perform a full inspection at the next available opportunity. The inspectors determined that the performance deficiency was more than minor because the licensees failure to inspect the fire barrier penetrations could result in not identifying degraded seals which could affect their ability to prevent a fire from spreading from one fire area to another. The finding was of very low safety significance because the failure to inspect a portion of fire barrier penetration seals did not impact the plants ability to reach and maintain safe shutdown. The finding has a cross-cutting aspect in the area of Human Performance, Work Management because the licensee improperly closed a notification to track the inspection of fire barrier penetrations without creating a work order. |