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| Start date | Site | Identified by | Title | Description | |
|---|---|---|---|---|---|
| 05000387/FIN-2018011-01 | 30 September 2018 23:59:59 | Susquehanna | NRC identified | Failure to conduct proper testing of 125 VDC molded case circuit breakers to confirm their design adequacy long-term | The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XI, Test Control. Specifically, Susquehanna has not established a program to adequately exercise and test safety-related 125VDC molded case circuit breakers (MCCBs) since initial plant operation. |
| 05000387/FIN-2018002-04 | 30 June 2018 23:59:59 | Susquehanna | NRC identified | EGM on Dispositioning BWR Licensee Noncompliance With TS Containment Requirements During Operations With A Potential For Draining The Reactor Vessel (EGM-11-03) | From April 2 through April 24, 2018, Susquehanna performed OPDRVs without establishing secondary containment integrity. 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. TS 3.6.4.1, Secondary Containment, requires that secondary containment be operable, and is applicable during OPDRVs. The required action for this specification if secondary containment is inoperable in this condition of applicability is to initiate actions to suspend OPDRVs immediately. As reported in LER 05000387/2018-001, Susquehanna conducted the following OPDRVs during the period of secondary containment inoperability: Recirculation system maintenance and pump replacement; Reactor water cleanup system flushes and maintenance; RHR system maintenance; Hydraulic control unit and control rod drive system maintenance; Local power range monitor replacements, including Intermediate Range Monitor 1E Dry Tube replacement; Control rod drive mechanism replacements; and Core spray instrument line flush. NRC EGM 11-03, EGM on Dispositioning BWR Licensee Noncompliance With TS Containment Requirements During Operations With A Potential For Draining The Reactor Vessel, Revision 3, provides, in part, for the exercise of enforcement discretion only if the licensee demonstrates that it has met specific criteria during an OPDRV activity. The inspectors assessed that Susquehanna adequately implemented these criteria. In accordance with EGM 11-003, in order to continue to receive enforcement discretion, a license amendment request (LAR) must be submitted and accepted for review within 12 months of the NRC staffs publication of the generic change, which occurred on December 20, 2016. The inspectors verified that Susquehanna submitted the required LAR on September 20, 2017 (ADAMS Accession No. ML17265A434), and that it was subsequently accepted by the NRC for review by a letter dated October 16, 2017 (ADAMS Accession No. ML17290A024).Corrective Action: Susquehanna submitted an LAR to adopt TS Task Force Traveler 542, Reactor Pressure Vessel Water Inventory Control, on September 20, 2017.Corrective Action Reference: AR-2015-01733 Enforcement: Violation: TS 3.6.4.1, Secondary Containment, requires that secondary containment be operable, and is applicable during OPDRVs. The required action for this specification if secondary containment is inoperable in this condition of applicability is to initiate actions to suspend OPDRVs immediately. Therefore, failing to maintain secondary containment operability during OPDRVs without initiating actions to suspend the operation was considered a condition prohibited by TSs as defined by 10 CFR 50.73(a)(2)(i)(B). Contrary to the above,from April 2 through April 24, 2018, Susquehanna performed OPDRVs without establishing secondary containment integrity. Basis for Discretion: The NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy because all criteria described in EGM 11-003 were met, enforcement discretion was previously authorized by EA-2017-089, and the licensee submitted an LAR on September 20, 2017 which was subsequently accepted by the NRC for review on October 16, 2017, and, therefore, will not issue enforcement action for this violation. The disposition of this violation closes LER 05000387/2018-001-00. |
| 05000387/FIN-2018002-01 | 30 June 2018 23:59:59 | Susquehanna | Self-revealing | Control Structure Chiller Inoperability Due to Identified Refrigerant Leaks Not Corrected | A Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action was self-revealed when the licensee failed to promptly correct a condition adverse to quality associated with the B control structure chiller which rendered the B control structure chiller inoperable. |
| 05000387/FIN-2018002-03 | 30 June 2018 23:59:59 | Susquehanna | NRC identified | Inadequate Justification for Deferral of Corrective Actions for certain Degraded Safety-Related Components | The inspectors identified a Green finding and associated NCV of TS 5.4.1, Procedures, when the licensee failed to promptly correct numerous operable but nonconforming or degraded safety-related components. |
| 05000387/FIN-2018002-02 | 30 June 2018 23:59:59 | Susquehanna | Self-revealing | Inadequate Procedure Adherence to Radiation Protection Requirements | A Green finding and associated NCV of Technical Specification (TS) 5.7, High Radiation Area, was self-revealed when two plant workers entered a posted high radiation area, and one workers electronic dosimeter alarmed on dose rate. The workers had not been briefed for entry into this area. |
| 05000387/FIN-2018410-02 | 31 March 2018 23:59:59 | Susquehanna | NRC identified | Security | |
| 05000387/FIN-2018410-01 | 31 March 2018 23:59:59 | Susquehanna | NRC identified | Security | |
| 05000387/FIN-2018010-01 | 31 March 2018 23:59:59 | Susquehanna | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section2.3.2 of the Enforcement Policy.Violation: 10 CFR 50.49(e)(5) requires, in part, that the electrical equipment qualification program must replace or refurbish the equipment at the end of its designated life.Contrary to the above, on November 16, 2017, the licensee identified that thirteen Unit 1, NAMCO limit switches in environmentally qualified (EQ) applications inside primary containment were not installed in their fully qualified configuration. Specifically, contrary to vendor instructions and EQAR-004 requirements, the limit switches for several containment isolation valves (CIV) have had their covers removed and reinstalled without replacing the gasket and cover screw O-rings. For this application, opening and/or removing the limit switch gasket /and cover screws O-ring constituted the end of the gasket/O-ring designated life. Significance/Severity Level: The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance (Green), because the limit switches provide only an open or closed signal indication in the main control room, so that operators are aware of the valve position, and can make appropriate assessment of plant conditions. The safety function of the containment isolation valves was not affected. |
| 05000387/FIN-2018001-01 | 31 March 2018 23:59:59 | Susquehanna | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Susquehanna Unit 1 TS section 5.4.1 requires that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Susquehannas implementing instruction NDAP-QA-0503, General Housekeeping, Transient Material and Internal Cleanliness, Revision 45 implements aspects of the Regulatory Guide administrative procedures requirements. NDAP-QA-0503 section 6.1.5.h requires, in part, that transient equipment shall be located such that it will not impact safety related equipment during a seismic event. Locate all items at a distance greater than the height of the item from safety related equipment. Additionally, TS 3.5.1 Action Statement I directs immediate entry into Limiting Condition for Operation (LCO)3.0.3 if one core spray subsystem is inoperable with one low pressure coolant injection (LPCI) subsystem inoperable. LCO 3.0.3 requires action to be taken within 1 hour to place the unit in MODE 2 within 7 hours and MODE 3 within 13 hours.Contrary to the above, from December 1, 2017 to December 3, 2017, Susquehanna staged a 540 pound, ten foot long replacement pipe on 34 inch high stands within 34 inches of the safety related Unit 1, B Core Spray room cooler. Susquehanna concluded that the room cooler was inoperable because the pipe could have reasonably contacted and damaged the flexible conduit for the power cable to the room cooler during a seismic event. Additionally, from 7:48 a.m. on December 2, 2017 to 1:35 p.m. on December 3, 2017, maintenance was performed on the Unit 1, division 2 LPCI swing bus motor generator which rendered the division 2 LPCI system inoperable. During this time, Susquehanna did not perform the required actions of LCO 3.0.3 and remained in MODE 1.Significance/Severity Level: This violation is of very low safety significance (Green), since this finding did not represent a loss of system, a loss of function of at least a single train for greater than its TS allowed outage time, or a loss of a non-TS train. Corrective Action Reference(s): CR-2017-20227; CR-2018-01717; CR-2018-02250 |
| 05000387/FIN-2017003-02 | 30 September 2017 23:59:59 | Susquehanna | NRC identified | RBCCW PCIV Design Control Issue | The inspectors identified a finding of very low safety significance (Green), an associated NCV of 10CFR50 Appendix B, Criterion III, Design Control, and a resultant violation of technical specification (TS) 3.6.1.3, Primary Containment Isolation Valves (PCIVs), when the reactor building closed cooling water (RBCCW) outboard isolation supply valve, HV21314, was found with a pull apart terminal block unseated within the motor control center (MCC), resulting in the loss of function for the valve to close given an initiation signal. Based on questions from inspectors, it was discovered that the terminal block was not installed in accordance w ith its dynamic qualification report. Immediate corrective actions included correctly seating the terminal block and performing an engineering evaluation to validate that the configuration conformed to the dynamic qualification report. The finding was more than minor because it was associated with the design control attribute of the reactor safety barrier integrity cornerstone and adversely 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 RBCCW outboard PCIV was inoperable for more than four years. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 3 of IMC 0609, Appendix A, The SDP for Findings At -Power, dated June 19, 2012, inspectors determined the significance to be of very low safety significance (Green) since the finding did not represent an actual open pathway in the containment isolation system and was not associated with hydrogen recombiners. The finding had a cross -cutting aspect in the area of Human Performance, Documentation, because Susquehanna did not maintain complete, accurate, and up -to-date documentation. Specifically, Susquehanna was not able to make a clear determination of the acceptability of the as -left configuration of the terminal block until the issue was discussed with the vendor to determine that the configuration was not in accordance with the dynamic qualification of the 480VAC MCC buckets. (H.7) |
| 05000387/FIN-2017405-01 | 30 September 2017 23:59:59 | Susquehanna | NRC identified | Security | |
| 05000387/FIN-2017003-01 | 30 September 2017 23:59:59 | Susquehanna | Self-revealing | Failure to Prepare Work Packages with Necessary Detail Results in Automatic Reactor Scram | The inspectors identified a self -revealing finding of very low safety significance (Green) because Susquehanna did not ensure that a work package was prepared to the detail necessary based on task difficulty in accordance with administrative procedure, NDAP- QA -0502, Revision 51. Specifically, on June 8, 2017, maintenance workers inadvertently shorted the Unit 1 main electro -hydraulic control (EHC) logic power supply to ground while working in a cabinet with little space to manipulate tools, resulting in a reactor scram. This finding is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, Susquehanna did not ensure measures were in place to prevent an adverse impact on the EHC control system during power supply voltage adjustment. This resulted in a rapid rise in reactor pressure and neutron flux, and subsequent automatic reactor scram. In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At -Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because while the performance deficiency caused a reactor scram, 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. The finding has a cross -cutting aspect in the area of Human Performance, Avoid Complacency, because the station failed to recognize and plan for the possibility of mistakes and inherent risk even while expecting successful outcomes. Specifically, individuals at various organizational levels did not ensure measures were in place to prepare maintenance technicians to perform a task on the EHC system that involved manipulating tools in a small space with tight clearances. |
| 05000387/FIN-2017002-03 | 30 June 2017 23:59:59 | Susquehanna | NRC identified | Follow -Up of Events and Notices of Enforcement Discretion | Inspection Scope For the plant event listed below, the inspectors reviewed and/or observed plant parameters, reviewed personnel performance, and evaluated performance of mitigating systems. The inspectors communicated the plant events to appropriate regional personnel, and compared the event details with criteria contained in IMC 0309, Reactive Inspection Decision Basis for Reactors, for consideration of potential reactive inspection activities. As applicable, the inspectors verified that Susquehanna made appropriate emergency classification assessments and properly reported the event in accordance with 10 CFR Parts 50.72. The inspectors reviewed Susquehannas follow - up actions related to the events to assure that Susquehanna implemented appropriate corrective actions commensurate with their safety significance. Unit 1, reactor scram due to transient initiated by an inadvertent loss of main turbine electrohydraulic control system control power due to a maintenance error . b. Findings No findings were identified. |
| 05000388/FIN-2017002-02 | 30 June 2017 23:59:59 | Susquehanna | NRC identified | Failure to Assess and Manage Risk Associated with Emergent Work | The inspectors identified a Green, self-revealing, NCV of 10 Code of Federal Regulations (CFR) 50.65 (a)(4) because Susquehanna failed to assess and manage the increase in risk for emergent work on the Unit 1 A 125 voltage direct current (VDC) battery charger. Susquehanna entered this issue into the CAP as CR-2017-09589. Corrective actions include conducting training on the emergent risk assessment process and reinforcing the expectation that control room staff is notified prior to releasing work. The PD was more than minor because it adversely impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of equipment performance involving availability and reliability. In addition, it is similar to Example 7.e from IMC 0612, Appendix E, Examples of Minor Issues, which states that the failure to perform an adequate risk assessment when required to do so is more than minor if the overall elevated plant risk would put the plant into a high licensee-established risk category and would require risk management actions under licensee procedures. The inspectors evaluated the significance using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management SDP and determined that this PD was of very low safety significance (Green). Specifically the PD was associated with risk management actions only and the incremental core damage probability (ICDP) was 2E-7 (<1E-6) for charger 1D613 out of service for approximately one hour.This finding had a cross-cutting aspect in the area of Human Performance, Consistent Process because individuals did not implement systematic approach to make decisions to commence work, and did not incorporate appropriate risk insights. (H.13) |
| 05000387/FIN-2017403-01 | 30 June 2017 23:59:59 | Susquehanna | NRC identified | Security | |
| 05000387/FIN-2017002-01 | 30 June 2017 23:59:59 | Susquehanna | NRC identified | Inadequate Assessment of Fire Brigade Performance during an Unannounced Drill | The inspectors identified a Green NCV of Susquehanna Unit 1 and 2 Operating License Condition 2.C.6, Fire Protection, because Susquehanna did not adequately assess an unannounced fire brigade drill, as required by the fire protection program. Susquehannaentered this issue into the corrective action program (CAP) for resolution as condition report(CR) CR-2017-10767 and is conducting an apparent cause evaluation to determine the most appropriate corrective actions.The performance deficiency (PD) was more than minor since the deficiency was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and impacted its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety Significance (Green) in accordance with D.1 of IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. Because the finding involved fire brigade training requirements, the fire brigade demonstrated the ability to meet the required times for fire extinguishment for the fire drill scenarios, and the finding did not significantly affect the fire brigades ability to respond to a fire, the finding screened as Green. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Self and Independent Assessments, because Susquehanna did not conduct assessments of their activities to assess performance and identify areas of improvement. Specifically, the Susquehanna self-evaluation of fire brigade performance was not of sufficient depth, appropriately objective, or self-critical. (P.6) |
| 05000387/FIN-2017001-01 | 31 March 2017 23:59:59 | Susquehanna | Self-revealing | Human Performance Error Results in Loss of Safety Secondary Containment Function | Green. A self-revealing finding of very low safety significance (Green) and associated NCV of TS 5.4.1, Procedures, was identified for failure to implement procedures that resulted in a secondary containment fan trip and associated loss of safety function. Susquehannas immediate corrective actions included restoring the secondary containment system to an operable configuration, and entering the issue into their corrective action program (CAP). Inspectors determined that the finding was more than minor because it was associated with the Human Performance attribute (Routine OPS/Maintenance Performance) of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (Secondary Containment) protect the public from radionuclide releases caused by accidents or events. The failure to adequately implement procedures for operation and maintenance of the secondary containment resulted in the inoperability of Zone 3 secondary containment and an associated loss of safety function. In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency only impacted the radiological barrier function of secondary containment. This finding had a cross-cutting aspect in the area of Human Performance, Teamwork because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, personnel did not conduct a re-brief of the team after the plan deviated from what was originally briefed, and the team did not adequately respond to challenges from workers in the field about whether it was appropriate to commence load center restoration with work still in progress. (H.4) |
| 05000387/FIN-2016004-01 | 31 December 2016 23:59:59 | Susquehanna | Self-revealing | Failure Rates Exceed (20%) Twenty Percent for Biennial Requalification Exam | Green. A self-revealing finding was identified associated with inadequate licensed operator performance during the annual licensed operator requalification operating test and biennial written examination. Specifically, 17 of 71 operators (23.9%) failed at least one portion of the requalification examinations. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of human performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, 17 of 71 licensed operators failed to demonstrate a satisfactory understanding of the required knowledge and abilities required to safely operate the facility under normal, abnormal, and emergency conditions. The inspectors evaluated this performance deficiency using IMC 0609, SDP, Appendix I, Licensed Operator Requalification SDP. This finding is of very low safety significance (Green) because the finding is related to requalification exam results, did not result in a failure rate of greater than 40 percent and all 17 operators were remediated and successfully retested prior to returning to licensed duties. This finding has a cross-cutting aspect in the area of Human Performance, Training, because Susquehanna did not provide adequate operator requalification training to maintain a knowledgeable, technically competent workforce. (H.7) |
| 05000387/FIN-2016007-01 | 31 December 2016 23:59:59 | Susquehanna | NRC identified | Failure to Specify and Maintain Safety-Related Quality Standards and Materials Essential for Reactor Core Isolation Cooling | The team identified a Green non-cited violation of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the failure to classify and maintain reactor core isolation cooling (RCIC) system components as safety-related as specified by Updated Final Safety Analysis Report Table 3.2-1 and Section 7.1.1. Specifically, although Talen, the operator of Susquehanna Steam Electric Station, classified the RCIC system as safety-related, this classification did not extend to the Unit 1 and Unit 2 RCIC barometric condenser relief valves. The team determined failure of the non-safety related barometric condenser relief valves could result in a loss of RCIC lube oil cooling and failure of RCIC to perform its design basis safety function. Talen entered the issue into the corrective action program as condition report 2016-23615 and performed an immediate operability determination, which concluded RCIC remained operable. The finding was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2 - Mitigating System Screening Questions. The team determined the finding screened as very low safety significance (Green), because the finding was a design deficiency which did not result in an actual loss of functionality of the RCIC system. This finding was not assigned a cross-cutting aspect because the performance deficiency occurred during original plant design and did not reflect current licensee performance. |
| 05000388/FIN-2016004-03 | 31 December 2016 23:59:59 | Susquehanna | Self-revealing | Refuel Floor Radiation Monitor Inoperable Due to being Improperly Calibrated | Green. A finding of very low safety significance (Green) and NCV of TS 5.4.1, Procedures was self-revealed when Susquehanna incorrectly calibrated the Unit 1 B refuel floor high exhaust duct high radiation monitor on November 15, 2014. This impacted the initiation capability of secondary containment isolation and control room emergency outside air supply system (CREOASS) and resulted in Susquehanna exceeding the allowed outage time for TSs 3.3.6.2, Secondary Containment Isolation, and 3.3.7.1, CREOASS Instrumentation. Upon identification of the issue, Susquehanna properly calibrated the radiation monitor to restore its operability. This finding is more than minor because it is associated with the Human Performance (Routine OPS/Maintenance Performance) attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (Secondary Containment and Control Room Ventilation) protect the public from radionuclide releases caused by accidents or events. Specifically, incorrectly calibrating the radiation monitor resulted in both systems being inoperable for almost two years. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The SDP for Findings At-Power, both dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was only associated with the radiological barrier function of the Control Room and Secondary Containment. This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency because Susquehanna did not recognize and plan for the possibility of mistakes, latent problems, or inherent risk, even while expecting successful outcomes. Specifically, Susquehanna personnel did not consider the potential undesired consequences of their actions before performing work and implement appropriate error-reduction tools (e.g. self-check, peer-check). (H.12) |
| 05000387/FIN-2016004-05 | 31 December 2016 23:59:59 | Susquehanna | NRC identified | LERs Associated with Reactor Coolant Pressure Boundary Leakage | Enforcement. TS 3.4.4, "RCS" requires RCS leakage be limited to no pressure boundary leakage in Mode 1. Contrary to this, pressure boundary leakage from a LPRM instrument housing and from socket weld #8 occurred between plant start-up in December 2015 and plant shutdown on June 6, 2016, and existed while in Mode 1. The inspectors determined that these violations of TS 3.4.4 are more than minor, but not the result of performance deficiencies. Specifically, for the first event, though leakage likely existed during the previous refueling outage when personnel were performing unrelated maintenance and inspection activities, it was likely too small to reasonably identify and correct. Similarly, for the 2016 leak identified in weld #8, the leakage causes were not within Susquehannas ability to foresee as they had replaced the weld with the industry recommended 2 x 1 taper configuration and used qualified procedures and personnel. The Susquehanna staff had also measured the susceptibility of the attached piping for vibrational inputs. In accordance with the NRC Enforcement Policy guidance and IMC 0612, these violations are being treated under the traditional enforcement process and best characterized as a Severity Level (SL) IV (very low safety significance) violation, similar to example d.1 in NRC Enforcement Policy, Section 6.1, Reactor Operations. Although a performance deficiency was not identified, to verify that the issue was of very low safety significance, the inspectors considered risk insights obtained by using IMC 0609, SDP, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that these TS violations would screen to Green (very low safety significance) because the boundary leakage would not have exceeded the leak rate for a small loss of coolant accident (LOCA) and would not affect any LOCA accident mitigating systems or components. Therefore, the inspectors considered that the SL IV characterization was appropriate. The licensee entered these issues into the Susquehannas CAP as CR-2016-14544 and CR-2016-14366. Because these issues are of very low safety significance, it has been determined that it was not reasonable for Susquehanna to be able to foresee and prevent, and as such no performance deficiencies exist. The NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of TS (EA-16-283). Further, because Susquehanna's actions did not contribute to this violation, it will not be considered in the assessment process or the NRC's Action Matrix. |
| 05000387/FIN-2016004-04 | 31 December 2016 23:59:59 | Susquehanna | Self-revealing | Auxiliary Bus Load Shed when a Daisy Chained Neutral was Interrupted during Maintenance | Green. A finding of very low safety significance (Green) for failure to develop an adequate work plan for replacement of a voltage potential indicating light on a breaker on the Unit 2 B auxiliary bus was self-revealed when the Unit 2 B reactor recirculation pump (RRP) tripped, along with other non-safety related loads on November 14, 2016, resulting in a rapid unplanned power change and transition to single loop operation. Specifically, operations and maintenance personnel did not recognize that disconnecting the neutral wires from the light socket would interrupt power to all of the degraded voltage relays for the auxiliary bus. Therefore, the relays de-energized when the maintenance was performed, tripping all the breakers on the bus. Susquehannas immediate corrective actions included stabilizing the plant, entering single loop operations, and entering the issue into their corrective action program (CAP). Additionally, Susquehanna performed a maintenance department stand down to communicate immediate lessons learned from the event while a more thorough causal analysis was conducted. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, implementation of work instructions resulted in the trip of the Unit 2 B RRP, B and D circulating water (CW) pumps, B and D condensate pumps, and the B service water (SW) pump, which caused an automatic trip of the C reactor feed pump and runback of the A RRP, resulting in a rapid power reduction to 32 percent rated thermal power (RTP). The inspectors evaluated the finding in accordance with IMC 0609, Appendix A "The SDP for Findings At-Power," dated June 19, 2012, Exhibit 1 for the Initiating Events cornerstone and determined the finding was of very low safety significance (Green) because it did not cause a reactor trip. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Management because Susquehanna did not implement a process of planning work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate with the work. Specifically, Susquehanna did not recognize the risk of interrupting a daisy chained neutral when planning a minor maintenance work order and did not recognize the impact of the work activity in the field. (H.5) |
| 05000387/FIN-2016004-02 | 31 December 2016 23:59:59 | Susquehanna | Self-revealing | Failure to Promptly Correct a Condition Adverse to Quality with LPCI Swing Bus Automatic Transfer Switches | Green. A finding of very low safety significance (Green) and associated NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed when Susquehanna failed to assure that conditions adverse to quality were promptly identified and corrected on two separate occasions. Both examples resulted in the failures of safety-related automatic transfer switches (ATSs) associated with the low pressure coolant injection (LPCI) swing buses. Corrective actions included enhancing the work instructions for all applicable ATSs based off original equipment manufacturer (OEM) input and scheduling the enhanced work instructions to be performed on the four swing bus ATSs during their next scheduled bus outages. Inspectors determined that the finding was more than minor because it was associated with the Equipment Performance attribute of the Reactor Safety Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In both examples, the failure to correct conditions adverse to quality resulted in the loss of power to the LPCI swing bus and inoperability of the respective division of LPCI. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, inspectors and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green). Specifically, though a single train was inoperable for greater than its technical specification (TS) allowed outage time, in consultation with regional senior reactor analysts, inspectors determined it did not represent an actual loss of function. The finding is related to the cross-cutting area of Problem Identification and Resolution, Evaluation, because Susquehanna did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Susquehanna either failed to evaluate deficiencies encountered during maintenance or failed to ensure that corrective actions aligned with and corrected the identified causes. (P.2) |
| 05000387/FIN-2016007-02 | 31 December 2016 23:59:59 | Susquehanna | Licensee-identified | Licensee-Identified Violation | Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion XVI, Corrective Actions, requires measures to be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to 10 CFR 50, Appendix B, Criterion XVI, station personnel did not promptly correct a condition adverse to quality. Specifically, from August 1, 2013 to February 26, 2016, Talen did not implement corrective actions to establish a PM program for molded case circuit breakers (MCCBs) found in distribution control panels that protect containment penetration conductors. The need for a MCCB PM program was originally identified during the 2013 NRC CDBI and documented as NCV 05000387 (05000388)/2013010-01, Failure to Verify Operation of Safety-Related 125Vdc Molded Case Circuit Breakers (CR 1732454). Talen identified this untimely implementation of corrective action during a self-assessment in preparation for the 2016 NRC CDBI. Plant staff entered the issue into the corrective action program (CRs 2016-04833; 23373; 23971 and 24015) and established a MCCB PM program. The team evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The team 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 (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. |
| 05000387/FIN-2016008-02 | 30 September 2016 23:59:59 | Susquehanna | NRC identified | Failure to Implement and Maintain Quality Procedure Results in Control Room Chiller Inoperability | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to implement and maintain a quality procedure, MT-GE-021, Chiller Maintenance and Inspection. This resulted in the safety related 0K112A chiller being operated outside of its design specifications and being declared inoperable. Specifically, on January 9, 2014, a system engineer directed the maintenance personnel to overcharge 0K112A with R-114a refrigerant, which led to higher power consumption by the chillers compressor motor, and the failure of the next biennial surveillance test on December 10, 2015 due to excessive compressor motor current. Susquehanna entered the issue into the CAP, conducted testing to establish the proper refrigerant charge, removed the excess refrigerant, and revised the procedure. The finding was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of Equipment Performance and adversely affected the associated cornerstone objective to ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The refrigerant overcharge condition resulted in the 0K112A chiller being inoperable and unable to fulfil its safety function to cool safety related switchgear and equipment during accident conditions for a period of 23 months. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined a detailed risk evaluation would be required because the finding involved an actual loss of function of at least a single Train for greater than its Technical Specification allowed outage time of 30 days. A detailed risk assessment was performed by a Region 1 Senior Reactor Analyst (SRA). The SRA determined the finding to be of very low safety significance (Green.) This finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence because individuals did not follow processes, procedures, and work instructions. Specifically, for many years maintenance and engineering personnel relied upon informal work practices vice referring to the procedure when charging the chillers with refrigerant. (H.8) (Section 4OA2.1.c(2)) |
| 05000388/FIN-2016003-01 | 30 September 2016 23:59:59 | Susquehanna | NRC identified | Inadequate Work Instructions for Breaching Internal Flood Barrier | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Susquehanna did not ensure that work instructions to breach a flood barrier appropriately incorporated design requirements for internal flooding so that equipment necessary to achieve and maintain safe shutdown would not be impacted. From August 30, 2016 to September 2, 2016, work instructions directed a breach of a flood barrier that was credited to provide assurance that equipment necessary for safe shutdown of the plant was protected against the effects of medium energy line breaks and, therefore, were not appropriate to the circumstances. Susquehanna entered this issue into their corrective action program (CAP) as condition report CR-2016-20472 and CR-2016-20859 and revised the work instructions to require a worker remain in the vicinity of the penetration to ensure that flooding could be secured prior to impacting equipment necessary to reach and maintain safe shutdown. This finding is more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, had the breach been completed, it could have allowed a medium energy line break in one flooding area to communicate with another area, potentially impacting equipment necessary to achieve and maintain safe shutdown. The inspectors evaluated the finding using IMC 0609, Appendix A, Exhibit 2, "Mitigating System Screening Questions," and determined the finding to be of very low safety significance (Green) because the PD was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of a safety function of a single train for greater than its technical specification allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect of Human Performance, Work Management because Susquehanna did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Implementation of Susquehannas work planning process did not ensure that the maintenance incorporated all requirements for protection against internal flooding and did not ensure that job site conditions were consistent with assumptions in engineering analyses. |
| 05000387/FIN-2016003-02 | 30 September 2016 23:59:59 | Susquehanna | NRC identified | Risk Management Actions Not Adequately Implemented | The inspectors identified a Green NCV of 10 CFR 50.65(a)(4) because Susquehanna did not assess and manage the increase in risk from online maintenance activities. From September 11 to 16, 2016, there were multiple affected areas that the fire protection engineer or designee did not walk down to inspect for fire impairments resulting in deficiencies not being corrected prior to releasing work and no fire watch was established for the impairments. Susquehanna removed the combustible materials from the areas or stationed a fire watch, and entered these issues into their CAP as CR-2016-21125, CR-2016-21423, CR-2016-21616, and CR-2016-21741. This finding is more than minor because it adversely impacted the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, not implementing the required risk management actions (RMAs) for the only available safe shutdown pathway placed the station in a much higher risk condition in the event of an internal fire. The inspectors evaluated the finding in accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. Since the performance deficiency was related to maintenance activities affecting structure, system, and components needed for fire mitigation, Appendix K directed the significance to be determined by an internal NRC management review using risk insights. IMC 0609, Appendix F, Attachment 1 Fire Protection Significance Determination Process Phase 1 Worksheet, was used to develop this risk insight. Based on the nature and quantity of combustible materials in the areas, combined with the relatively short duration of which the fire risk was unmitigated, inspectors determined that it was of very low safety significance (Green). The finding was determined to have a cross-cutting aspect in the area of Human Performance, Avoid Complacency, in that, individuals did not plan for latent issues and inherent risk, even while expecting successful outcomes. Specifically, combustible materials were not appropriately controlled as required by OI-013-002, Fire Risk Management, Revision 10, because in some cases they were assumed to be exempt from the program requirement or staff did not tour the areas because they assumed there were no combustible materials present based on past experience. |
| 05000387/FIN-2016406-01 | 30 September 2016 23:59:59 | Susquehanna | NRC identified | Security | |
| 05000387/FIN-2016008-04 | 30 September 2016 23:59:59 | Susquehanna | Self-revealing | Failure to Promptly Identify and Correct a Condition Adverse to Quality on Vital 480 VAC MCCs | The inspectors documented a self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to identify and correct a condition adverse to quality. Specifically, in October and December 2006 and July 2009, Susquehanna did not identify a non-conforming condition with the design and performance requirements of several 480 volt motor control center (MCC) breaker assemblies during receipt inspections. These non-conforming breaker assemblies were installed in vital 480 VAC applications and subsequently led to a phase to ground short and loss of a 480 volt safety-related motor control center on May 12, 2016. Susquehanna entered this issue into their CAP, conducted an apparent cause evaluation, replaced the damaged breaker assembly, and is conducting an extent of cause review for other susceptible breaker assemblies. The finding was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, on May 12, 2016, an electrical transient on vital AC bus 2B246 occurred as a result of a phase to ground fault in breaker cubicle 2B24609, which resulted in a loss of bus 2B246 and associated safety related loads. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not cause both a reactor trip and loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding did not have a crosscutting aspect because the performance deficiency was a historical issue with the actions taken in 2005, 2006, and 2009, and is not indicative of current licensee performance. (Section 4OA2.1.c(4)) |
| 05000388/FIN-2016008-03 | 30 September 2016 23:59:59 | Susquehanna | Self-revealing | Failure to Implement or Develop Timely Interim or Final Corrective Actions for a Degraded Condition | The inspectors documented a self-revealing finding of very low safety significance (Green) against Susquehanna procedures LS-125 Revision 4, Corrective Action Program (CAP), and OI-AD-096 Revision 18, Operator Challenges, for the failure to correct and establish appropriate corrective actions for a known degraded condition for an uninterruptable power supply (UPS) for vital 120 VAC load centers. Specifically, Susquehanna did not correct nor establish compensatory actions for the transfer switch for a UPS which was failed for over one year. The degraded condition subsequently complicated operator response to the loss of a vital 480 VAC switchboard and resulted in an unplanned manual reactor scram and valid emergency core cooling system (ECCS) actuation on May 13, 2016. Susquehanna entered this issue into their CAP, conducted an apparent cause evaluation, and repaired the UPS transfer switch. The finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the long standing degraded condition of UPS 2D14212/2B246082 was not corrected or compensated for and did not function as designed, as a result operators had to manually scram the reactor following the loss of a vital bus on May 13, 2016. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not cause both a reactor trip and loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, while this performance deficiency resulted in a reactor scram, it was not the cause of the loss of mitigation equipment credited in the Susquehanna safety analysis. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution Resolution because the organization did not take effective corrective actions to address issues in a timely manner commensurate with its safety significance. Specifically, failing to establish appropriate compensatory actions for this known degraded condition, prevented the operators from responding appropriately to a loss of a vital 480 VAC switchboard initiating event. (P.3). (Section 4OA2.1.c(3)) |
| 05000387/FIN-2016008-01 | 30 September 2016 23:59:59 | Susquehanna | NRC identified | Failure to Write a Condition Report for Degraded Conditions Which Challenged Operability of Safety Related Equipment | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Susquehanna failing to identify and correct conditions adverse to quality in a timely manner. Specifically, between April 16, 2016 and April 22, 2016, condition reports for potential or suspected degraded or non-conforming conditions related to the High Pressure Coolant Injection System (HPCI) and Reactor Core Isolation Cooling System (RCIC) were not written and operability determinations performed. In both cases, the equipment was subsequently declared inoperable due to the conditions. The issues were entered into the CAP and the equipment was taken out of service, repaired, and retested satisfactorily. The inspectors determined that there were two examples of the same performance deficiency and violation. In accordance with NRC Enforcement Manual Section 1.3.4, Documenting Multiple Examples of a Violation, multiple examples of a single violation are allowed to be documented as a single violation bounded by the characterization of the most significant example. The RCIC example is considered the most significant due to the longer exposure time in a required mode and number of mode changes that occurred during the exposure period. The finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to identify and correct degraded conditions associated with a RCIC system lube oil leak which rendered that system inoperable. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that this finding screened to Green because the safety function was not lost, and the finding did not represent an actual loss of function of at least a single train for greater than its Tech Spec Allowed Outage Time or two separate safety systems out-of-service for greater than its Tech Spec Allowed Outage Time. This finding had a cross-cutting aspect in the area of Human Performance, Teamwork, because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, in both examples, individuals were aware of potential degraded conditions but actions were not taken to communicate the activity to other groups, such as the control room operators, to allow for the issues to be evaluated for operability and determine if proposed actions were timely and/or appropriate. (H.4) (Section 4OA2.1.c(1)) |
| 05000387/FIN-2016002-06 | 30 June 2016 23:59:59 | Susquehanna | Self-revealing | Entry into a Locked High Radiation Area without Radiological Briefing | A Green self-revealing NCV of TS 5.7.2, High Radiation Area Controls, was identified when workers entered the wrong reactor unit condenser bay (Unit 2) that was posted and controlled as a locked high radiation area (LHRA). Specifically, on May 3, 2016, four Susquehanna staff were briefed to enter the Unit 1 condenser bay to check for steam leaks during start up, however the staff entered the Unit 2 condenser bay during full power operations in error and received electronic dosimeter alarms. This was entered into the CAP as CR-2016-11944, the use of master keys for routine entry into LHRA was discontinued, and a radiation safety stand down was conducted. The finding was determined to be more than minor based on a similar example 6.h in IMC 0612, Appendix E, and it is associated with Human Performance attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, Susquehanna staff violated the RWP and briefing requirements designed to protect workers from unnecessary radiation exposure. Using IMC 0609, Appendix C, Occupational Radiation Safety SDP, dated, August 19, 2008, the finding was determined to be of very low safety significance (Green) because it did not involve: (1) ALARA occupational collective exposure planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The finding was self-revealing because Susquehanna was made aware of the situation as a result of an electronic dose rate alarm. The finding is related to the cross-cutting area of Human Performance, Teamwork because the workers did not conduct peer checking and recognize and communicate that they were in the wrong reactor unit for the work they were conducting. Specifically, four Susquehanna staff were briefed to enter the Unit 1 condenser bay to check for steam leaks during start up, however the staff entered the Unit 2 condenser bay. |
| 05000387/FIN-2016002-02 | 30 June 2016 23:59:59 | Susquehanna | Self-revealing | Failure to Promptly Correct a Condition Adverse to Quality with A EDG MOC Switch | A self-revealing finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to correct a condition adverse to quality. Specifically, on March 23, 2016, the A emergency diesel generator (EDG) failed its technical specification (TS) surveillance test in that the emergency switchgear room cooler, 1V222A, started immediately when the EDG loaded onto the emergency bus following a simulated loss of off-site power (LOOP) and simulated Emergency Core Cooling System (ECCS) Initiation, rather than sequencing onto the bus as intended by design. Susquehanna identified the direct cause of the failure was due to a misadjustment of the mechanism-operated cell (MOC) linkage switch (S1) in the A EDG output breaker to the 1A 4 kilovolt (kV) bus, which provides the electrical logic to the 1V222A load timer. The repeat failure was entered into the corrective action program (CAP) as CR-2016-08643, the MOC linkage was realigned, and the functions satisfactorily tested. The finding was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to correct the degraded condition rendered the A EDG inoperable for longer than the TS allowed outage time. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding required a detailed risk assessment because the finding represents an actual loss of function of a single train for greater than the TS allowed outage time. Specifically, the A EDG was inoperable from July 19, 2010 until April 2, 2016, because TS requires functioning of the sequencing timers for the EDG to be operable. In coordination with a Region 1 Senior Risk Analyst, the issue was qualitatively screened as Green (very low safety significance) based on the low initiating event frequency associated with a loss of coolant accident (LOCA) co-incident with a LOOP event, and observed successful EDG function during multiple LOOP/LOCA tests over the period in question. This would result in a delta core damage frequency substantially less than E-6. Additionally, it was reasonable to conclude that the A EDG remained available to perform its function given the minimal increased load on the machine as evidenced during the performance of the LOOP-LOCA surveillance testing in 2012, 2014, and 2016. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Susquehanna did not thoroughly evaluate the issue to ensure that the resolution addressed the cause and extent of conditions commensurate with their safety significance. Specifically, Susquehanna corrected a suspected condition without appropriate troubleshooting until the third identical failure of the 1V222A load timer. |
| 05000387/FIN-2016002-03 | 30 June 2016 23:59:59 | Susquehanna | Self-revealing | Failure of B EDG to Reach Rated Frequency within 10 Seconds | A self-revealing finding of very low safety significance (Green) and associated NCV of TS 5.4.1.a, Procedures, was identified when Susquehanna failed to implement procedures for loading EDGs promptly following extended unloaded operation. Specifically, Susquehanna did not load the B EDG promptly following over 6 hours of unloaded operation which resulted in the slow starting time during the subsequent surveillance test due to insufficient fuel delivery caused by clogged fuel injectors. The failure was entered into the CAP as CR-2016-13220 and the EDG was run loaded for an extended period to ensure any unburned fuel had been removed from the machine. The finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and affected the objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the failure to load the B EDG following extended operation unloaded resulted in the slow starting time of the EDG during subsequent surveillance testing due to clogged fuel injectors. The inspectors evaluated the finding in accordance with Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012 and determined that it was of very low safety significance (Green) because it did not affect the design or qualification of the EDG, did not represent a loss of system function, and did not represent a loss of a single train for greater than its TS allowed outage time. The finding is related to the cross-cutting area of Human Performance, Consistent Process, because Susquehanna did not use a consistent, systematic approach to make decisions which incorporated risk insights. Specifically, Susquehanna did not appropriately coordinate the loaded run of the B EDG with maintenance on the C EDG to ensure B EDG availability was not unnecessarily challenged. |
| 05000387/FIN-2016002-04 | 30 June 2016 23:59:59 | Susquehanna | NRC identified | Failure to Critique an Incorrect PAR Notification | An NRC-identified finding of very low safety significance (Green) and associated NCV of 10 CFR 50.54(q)(2),Emergency Plans was identified when Susquehanna failed to identify that an incorrect notification of wind direction was made to the senior state official (SSO) during a full-scale drill. This failure was entered into the CAP as CRs 2016-14303 and 2016-14128, ERO personnel involved in the incorrect communication and the drill controllers that failed to identify the deficiency were remediated, and lessons learned communicated to other emergency response organization personnel. The finding was more than minor because it is associated with the emergency response organization (ERO) Performance attribute of the Emergency Preparedness Cornerstone and affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the failure of Susquehanna personnel to effectively identify an exercise weakness associated with a risk significant planning standard (RSPS) caused a missed opportunity to identify and correct a drill-related performance deficiency. The inspectors assessed the issue using the Emergency Preparedness SDP, Appendix B to IMC 0609, dated September 23, 2014. Susquehanna's failure to critique the inaccurate notification met the NRC's definition of a weakness in a full-scale drill. However, because four previous notifications had accurately reported the wind direction and the miscommunication was inconsistent with the correct protective actions recommendation (PAR) that was communicated simultaneously, in consultation with a senior emergency preparedness inspector, inspectors determined the communication would likely have been corrected prior to the offsite response organizations (OROs) acting on the incorrect information, did not result in an incorrect PAR, and therefore determined that that the failure to critique the drill weakness only constituted a degradation of the planning standard (PS) function. Therefore the finding is characterized as having very low safety significance (Green). The finding is related to the cross-cutting area of Problem Identification and Resolution, Identification, in that Susquehanna did not identify a RSPS issue completely, accurately, and in a timely manner commensurate with the safety significance. Specifically, during the full-scale drill, Susquehanna failed to recognize and critique that a RSPS was not met and did not place this issue into the CAP until prompted by inspectors. |
| 05000387/FIN-2016002-05 | 30 June 2016 23:59:59 | Susquehanna | Self-revealing | Entry into a High Radiation Area without Radiological Briefing | A Green self-revealing NCV of TS 5.7.1, High Radiation Area Controls, was identified when a worker did not comply with a radiological posting barrier and other access control requirements for high radiation area (HRA) entry. Specifically, on December 26, 2015, a security officer entered into a posted HRA without proper authorization. This was entered into the CAP as CR-2015-33947, the HRA barrier was moved further out, and a shield rack was placed in front of the condenser bay door to reduce radiation dose rates. The finding was determined to be more than minor based on similarity to example 6.h in IMC 0612, Appendix E, and it is associated with Human Performance attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the individual violated the HRA posting, radiation work permit (RWP) and briefing requirements designed to protect the worker from unnecessary radiation exposure. Using IMC 0609, Appendix C, Occupational Radiation Safety SDP, dated August 19, 2008, the finding was determined to be of very low safety significance (Green) because it did not involve: (1) as low as is reasonably achievable (ALARA) occupational collective exposure planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The finding is related to the cross-cutting area of Problem Identification and Resolution, Resolution, in that the organization did not ensure that corrective actions to address the cause of repetitive electronic dosimeter alarms in this area of the plant and had not been sufficiently evaluated and had not enhanced radiological controls to prevent this issue from recurring. |
| 05000387/FIN-2016002-01 | 30 June 2016 23:59:59 | Susquehanna | Self-revealing | Failure to Promptly Identify a Condition Adverse to Quality Associated with Primary Containment Isolation Valves | A self-revealing Green finding and associated violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, and TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), was identified when Susquehanna did not promptly identify a condition adverse to quality. Despite observing abnormal behavior during local leak rate testing following replacement in May 2014, Susquehanna did not take any action to ensure that certain Reactor Water Cleanup (RWCU) system PCIVs passed their subsequent testing. Consequently, these valves failed their in-service and local leak rate test in March 2016 when they failed to close upon securing system flow. The failure was caused by an internal interference between the check valve hinge and body. Following the failures in March 2016, Susquehanna repaired the valves and successfully performed local leak rate testing, restoring operability of the PCIVs. The repeat failure was entered into the CAP as CRs 2016-06960 and 2016-09940. The finding was determined to be more than minor because it was associated with the Structure, System, and Component (SSC) and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to identify a condition adverse to quality during post-maintenance testing resulted in two PCIVs being rendered inoperable for longer than the TS allowed outage time. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not involve the hydrogen recombiners and did not result in an actual open pathway in the physical integrity of reactor containment. Specifically, the redundant valve for each penetration remained operable during the period in which these two valves were inoperable. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because Susquehanna did not use decision making practices that emphasized prudent choices over those that are simply allowable. Specifically, Susquehanna decided to accept elevated seat leakage for two new PCIVs, assuming that they could be declassified as PCIVs. |
| 05000388/FIN-2016002-07 | 30 June 2016 23:59:59 | Susquehanna | NRC identified | HPCI Overridden Prior to Manual Reactor Scram | An NRC-identified finding of very low safety significance (Green) and associated NCV of TS 5.4.1.a, Procedures, was identified when Susquehanna failed to implement procedures for controlling the high pressure coolant injection (HPCI) system. Specifically, operators overrode automatic initiation of the system prior to inserting a manual scram, contrary to the requirements of OP-252-001, HPCI System, and OP-AD-300, Administration of Operations. This was entered into the CAP as CRs 2016-12854 and 2016-13118 and 2016-13136, the operators involved in the event were remediated, and lessons learned communicated to other station personnel. The finding was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems Cornerstone and affected the objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, overriding the HPCI system prior to initiating a plant scram rendered the system unavailable to respond to a level transient or failure of the non-safety related feedwater system. The inspectors evaluated the finding in accordance with Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012 and determined that it required a detailed risk assessment because it represented a loss of the single train systems function. The Region 1 SRA performed a detailed risk evaluation using the Susquehanna Unit 2 standardized plant analysis risk (SPAR) Model, version 8.23. The issue was conservatively modeled with a HPCI failure to start due to the system automatic start signal being overridden. The change in core damage frequency per year was determined to be in the E-10 range due to the very short duration the system auto start feature was defeated. Therefore the issue was determined to be of very low safety significance (Green). The finding is related to the cross-cutting area of Human Performance, Procedure Adherence because Susquehanna did not follow processes, procedures and work instructions. Specifically, operators did not ensure that their actions were appropriately authorized by procedures when taking action to override a key safety system prior to a plant transient. |
| 05000388/FIN-2016001-05 | 31 March 2016 23:59:59 | Susquehanna | Licensee-identified | Licensee-Identified Violation | 10 CFR Part 50 Appendix B, Criterion III Design Control, requires, in part, that measures be established for the selection and review for suitability of applicable equipment that is essential to the safety-related functions of systems. Susquehanna Unit 2 TS 3.3.5.1 requires operability of four reactor steam dome pressure low channels to provide the injection permissive for CS (Function 1d) and LPCI system (Function 2d). Contrary to this requirement, Susquehanna did not consider the effects of mechanical hysteresis on the function of reactor steam dome pressure switches when operated above their normal operating range. These reactor steam dome pressure switches remain in an overpressure condition during normal operations and as reported in LER 388/2015-001 resulted in pressure switches drifting out of TS allowable values for emergency core cooling system (ECCS) injection valve permissive interlocks for greater than the TS allowed outage time. This violation has been entered into Susquehannas CAP as CR-2015-06243. In consultation with regional Senior Reactor Analysts, the inspectors determined the finding was of very low safety significance (Green) because the ability to open low pressure ECCS injection valves manually remained available and engineering analysis for the as-found condition of the switches determined that the resultant delay in automatic response would have a negligible increase in peak central temperature during a design basis accident. |
| 05000387/FIN-2016001-02 | 31 March 2016 23:59:59 | Susquehanna | NRC identified | Failure to Report Loss of Safety Function as Required by 10 CFR 50.73(a)(2)(v) | Inspectors identified a Severity Level IV NCV of 10 CFR Part 50.73 (a)(2)(v) when Susquehanna did not submit a licensee event report (LER) within 60 days of identifying that both trains of the control room emergency outside air supply system (CREOASS) were rendered inoperable during surveillance testing, a condition that could have prevented fulfillment of a safety function. Susquehanna entered the issue into the CAP as CR-2016-03713 and reported the condition on May 5, 2016 in LER 50-388(387)/2015-015. Since the issue had the potential to affect the NRCs ability to perform its regulatory function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that it was a Severity Level IV violation. The significance of the associated performance deficiency was also screened against the reactor oversight process (ROP) per the guidance of IMC 0612, Appendix B, "lssue Screening. Because this violation involves the traditional enforcement process and does not have an associated finding under the ROP, inspectors did not assign a cross-cutting aspect to this violation. |
| 05000388/FIN-2016001-01 | 31 March 2016 23:59:59 | Susquehanna | NRC identified | Failure to Assess and Manage Risk of Maintenance Activities for a SLC System Flow Surveillance | The inspectors identified a Green NCV of 10 CFR 50.65(a)(4) because Susquehanna did not adequately assess the risk of performing maintenance in accordance with station procedures. Specifically, Susquehanna did not assess the risk of performing a standby liquid control (SLC) system flow surveillance in conjunction with having the D emergency diesel generator (EDG) unavailable and therefore did not specify appropriate risk management actions (RMAs). Susquehanna entered the issue into the CAP as CR-2016-04137. The inspectors determined that this performance deficiency is more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, the finding is similar to example 7.e. in NRC IMC 0612 Appendix E, Examples of Minor Issues. This example states, in part, that failure to perform an adequate risk assessment when required by 10 CFR 50.65 (a)(4) is not minor if the overall elevated plant risk would put the plant into a higher licensee-established risk category or would require, under plant procedures, RMAs or additional RMAs. In this case, the combination of the D EDG maintenance and SLC flow surveillance resulted in changing risk to Yellow which required additional RMAs in accordance with station procedures. The inspectors evaluated the finding using IMC 0609 Appendix K, Maintenance Risk Assessment and Risk Management SDP. The inspectors and the Region I senior resident analyst used Appendix K, Flowchart 1, Assessment of Risk Deficit, and determined that the inadequate risk assessment was of very low safety significance (Green). The basis for this determination was that the short duration of the actual planned maintenance activities (3.5 hours) associated with the SLC unavailability results in less than E-9 calculated incremental core damage probability deficit (ICDPD) using Susquehannas risk model. Since the resultant ICDPD is below 1 E-8 threshold, the finding was determined to be Green. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Management in that Susquehanna did not appropriately incorporate insights from probabilistic risk assessments into the daily work activities (H.5). Specifically, Susquehanna did not appropriately assess the risk of performing maintenance activities as specified in station procedures. |
| 05000387/FIN-2016001-03 | 31 March 2016 23:59:59 | Susquehanna | Self-revealing | Failure to Correct Fatigue Related Cracking of the B RRP Lower Seal Cavity Vent Line | A self-revealing finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for Susquehannas failure to establish measures to assure a condition adverse to quality was corrected. Specifically, vibration induced fatigue cracking on the Unit 1 B reactor recirculation pump (RRP) lower seal cavity vent piping was not corrected in December 2014 after a reactor coolant pressure boundary leak had occurred. This resulted in another reactor coolant pressure boundary leakage at the same location with Unit 1 operating in Mode 1, a condition prohibited by technical specifications (TS) LCO 3.4.4. Susquehannas entered the issue into the corrective action program (CAP) as CR-2015-30901 and replaced and modified the union that included the weld. The finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding in accordance with Exhibit 1 of IMC 0609, Appendix A, The significance determination process (SDP) for Findings At-Power, and determined the finding was of very low safety significance (Green) because the leakage would not have exceeded the reactor coolant system (RCS) leak rate for a small loss of coolant accident (LOCA) and it did not affect other systems used to mitigate a LOCA. This finding had a cross-cutting aspect in the area of Human Performance, Work Management, because Susquehanna did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, in that Susquehanna did not adequately coordinate the work activities with different groups (H.5). Specifically, welding engineers were not engaged in the decision making process during the December 2014 repair and consequently the repair was inadequate to ensure the entire crack had been removed. |
| 05000387/FIN-2016001-04 | 31 March 2016 23:59:59 | Susquehanna | Licensee-identified | Licensee-Identified Violation | 10 CFR Part 50 Appendix B, Criterion XVI Corrective Action, requires, in part, that measures be established to assure conditions adverse to quality are promptly identified and corrected. Susquehanna Unit 1 TS 3.3.8.2 requires operability of two RPS electrical power monitoring assemblies for each inservice RPS alternative power supply. Contrary to this, a defective electrical power monitoring assembly was installed into the RPS alternative power supply in December 2014. Susquehanna identified that in September 2000, an electrical power monitoring assembly had failed to perform an UF trip. An evaluation of the failure was conducted and Susquehanna incorrectly concluded in 2001 that the electrical power monitoring assembly did not contribute to the failure and returned the electrical power monitoring assembly back to a spare parts status. In December 2014 the subject electrical power monitoring assembly was installed into the alternative RPS power supply and in September 2015 again failed to perform an UF trip. As reported in LER 387/2015-008 this resulted in an inoperable electrical power monitoring assembly in the inservice alternative RPS power supply for greater than the allowed outage time. This violation has been entered into Susquehannas CAP as CR-2015-25881. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Because it was associated with the RPS, the inspectors screened the violation against the questions in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, Part C, Reactivity Control Systems. The inspectors determined the finding was of very low safety significance (Green) because the second RPS electrical power monitoring assembly remained operable which would have generated an UF trip for the RPS alternative power supply if required, did not involve unintentional reactivity addition, and did not involve a mismanagement of reactivity by operators. |
| 05000387/FIN-2015004-05 | 31 December 2015 23:59:59 | Susquehanna | Licensee-identified | Licensee-Identified Violation | On March 6, 2015, Susquehanna identified a programmatic deficiency associated with the perimeter radiation monitoring system (RMS), which consists of 16 fixed radiation monitors at the site boundary that are activated and monitored by TSC/EOF staff, in that it was in a condition that could impact the ability to assess the EAL thresholds. Specifically, Susquehanna identified that the RMS is not maintained as required by EP-115, Equipment Important to Emergency Response, Revision 2, in that 4 of the 16 fixed radiation monitors in the RMS had been out of service since 2013 and the software that displays the RMS does not consistently run in computers in the EOF and TSC. 10 CFR Part 50.54(q) requires that the facility licensee follow and maintain in effect emergency plans which meet the standards in 10 CFR 50.47(b). 10 CFR 50.47(b)(4) requires, in part, that emergency response plans include a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters. Susquehannas Emergency Plan identifies the RMS as an initiating condition for EALs RG1 and RS1. Specifically, the two thresholds are met if sustained readings on the RMS are above the specified threshold value for greater than 15 minutes. Contrary to the above, the RMS was not maintained as required by the Emergency Plan which could have impacted the ability of Susquehanna to declare an emergency event. Susquehanna entered this issue into the CAP as CR-2015-06706. The inspectors determined that the finding was 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 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 determined, through a review of IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, issued September 23, 2014, the finding to be of very low safety significance (Green), because redundant EAL initiating conditions associated with offsite dose assessment would have ensured that Susquehanna maintained a capability to declare the site area and general emergencies affected by the RMS. |
| 05000387/FIN-2015201-02 | 31 December 2015 23:59:59 | Susquehanna | NRC identified | Security | |
| 05000387/FIN-2015404-01 | 31 December 2015 23:59:59 | Susquehanna | NRC identified | Security | |
| 05000387/FIN-2016404-01 | 31 December 2015 23:59:59 | Susquehanna | NRC identified | Security | |
| 05000387/FIN-2015403-02 | 31 December 2015 23:59:59 | Susquehanna | Licensee-identified | Licensee-Identified Violation | |
| 05000387/FIN-2015004-04 | 31 December 2015 23:59:59 | Susquehanna | Self-revealing | Inadvertent Closure of the B Inboard MSIV | A self-revealing finding of very low safety significance (Green) was identified when Susquehanna did not correctly validate a deficient condition associated with the Unit 1 B inboard main steam isolation valve (MSIV) direct current (DC) solenoid valve as an actual valve issue, vice indication-only, through the use of specific acceptance criteria as required by MT-AD-509, Control of Minor Maintenance Activities. By incorrectly concluding the issue was indication only, testing was allowed to be performed which inserted a half-isolation by de-energizing the alternating current (AC) solenoid valve on the B inboard MSIV. When this maintenance was performed, the B inboard MSIV closed unexpectedly, resulting in a reactor scram. The cause of the closure was the failure of the DC solenoid valve on the B inboard MSIV. Susquehanna entered the issue into the CAP as CR-2015-30721 and replaced the DC solenoid for the B MSIV. The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the maintenance activity performed to validate the DC solenoid valve continuity was inadequate and as a result the testing was allowed to be performed which relied on DC solenoid valve continuity to prevent an MSIV closure. The inadvertent closure of the B inboard MSIV resulted in a high pressure scram. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, for the Initiating Events cornerstone. The inspectors determined the finding was of very low safety significance (Green) because it did not cause the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, the condenser was maintained for decay heat removal via the bypass valves through the other three main steam lines following the trip. This finding had a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Susquehanna did not stop when faced uncertain conditions and instead rationalized unanticipated test results. Specifically, the investigation of the extinguished continuity monitor focused on the possibility that it was an indication-only issue and failed to question the acceptability of the current values obtained during troubleshooting (H.11). |
| 05000388/FIN-2015004-01 | 31 December 2015 23:59:59 | Susquehanna | Self-revealing | Failure to Correct a Condition Adverse to Quality Associated with an Inoperable Primary Containment Isolation Valve | A self-revealing finding of very low safety significance (Green) and associated violations of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, and Technical Specification (TS) 3.6.1.3, Primary Containment Isolation Valves (PCIVs), was identified when Susquehanna did not take adequate corrective action to address the inoperability of the reactor recirculation sample line outboard PCIV when it failed during surveillance testing on July 1, 2015. The valve failed its subsequent surveillance test on September 30, 2015 due to the same degraded condition, which rendered the valve inoperable for longer than the allowed outage time specified in TS 3.6.1.3. The repeat failure was entered into the CAP as CR-2015-26590 and restored the valve to an operable condition by replacing its associated solenoid valve. The finding was determined to be more than minor because it was associated with the structure, system and component (SSC) and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to correct the degraded condition of solenoid valve sticking resulted in a PCIV being rendered inoperable for longer than the TS allowed outage time. Inspector evaluated the finding In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, and determined it is of very low safety significance (Green) because the performance deficiency did not result in an actual open pathway in the physical integrity of reactor containment, because the inboard valve remained operable for the duration of the inoperability, and it did not involve the hydrogen recombiners. This finding had a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Susquehanna did not stop when faced with uncertain conditions and ensure the risks were evaluated and managed before proceeding. Specifically, Susquehanna did not challenge the unanticipated test results and did not ensure that the condition adverse to quality, associated with the faulty solenoid valve, was resolved prior to considering the valve operable (H.11). |