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| Site | Quarter | Title | Description | |
|---|---|---|---|---|
| 05000331/FIN-2018003-02 | Duane Arnold | 2018Q3 | Minor Violation | During Mode 1 power operations on July 9, 2018, the licensee had both doors of a secondary containment airlock open simultaneously, and a minor violation of Technical Specification (TS) 3.6.4.1 Secondary Containment was self-revealed. During the time both doors were open, approximately 3 seconds, the allowable penetration opening area was exceeded and rendered the secondary containment inoperable. Technical Specification 3.6.4.1 requires secondary containment to be operable in Modes 1, 2 and 3. Technical Specification Surveillance Requirement 3.6.4.1.2 supports secondary containment operability by verifying that either the outer door(s) or the inner door(s) in each secondary containment access opening are closed. The posted instructions at each secondary containment airlock door stated, ATTENTION Push Button To Be Held In For 2 Seconds Prior To Opening Door, to be of a type appropriate for traversing the containment airlock. Contrary to the above, at approximately 1:34 p.m. on July 9, 2018, while operating in Mode 1 at 97 percent power, two individuals simultaneously traversing through opposite doors of a secondary containment airlock each failed to hold the airlock interlock push button for two seconds prior to opening their respective doors resulting in a momentarily inoperability of secondary containment. Operability was restored upon the immediate closure of one of the two doors. Subsequently, maintenance was unable to recreate the condition and satisfactorily performed Surveillance Test Procedure (STP) 3.6.4.102, Secondary Containment Airlock Verification, and GMPELEC44,Section A5.1,Airlock Door Interlock Checks.The licensee entered this |
| 05000354/FIN-2018003-04 | Hope Creek | 2018Q3 | Enforcement Action (EA)-18-044: EGM on Dispositioning BWR Licensee Noncompliance With TS Containment Requirements During Operations With A Potential For Draining The Reactor Vessel (EGM-11-003) | From April 19 through April 29, 2018, HCGS 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 3.6.5.1, Secondary Containment Integrity, requires that secondary containment integrity be maintained, and is applicable during OPDRVs. The required action for this specification without secondary containment integrity in this condition of applicability is to suspend OPDRVs. As reported in LER 05000354/2018-001, HCGS conducted the following OPDRVs during the period of secondary containment inoperability: Control rod drive mechanism replacements; Local power range monitor replacements; and Cavity let down via Reactor Water Clean Up system. Additionally, an unplanned OPDRV occurred due to RHR system relief valves seat leakage. 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 HCGS 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 that occurred on December 20, 2016. The inspectors verified that PSEG submitted the required LAR on September 20, 2017 (ADAMS Accession No. ML17265A847), and that it was subsequently accepted by the NRC for review by a letter dated October 25, 2017 (ADAMS Accession No. ML17299A009). Corrective Action: PSEG submitted an LAR to adopt TS Task Force Traveler 542, Reactor Pressure Vessel Water Inventory Control, on September 20, 2017, that was subsequently accepted by the NRC for review on October 25, 2017. (After the end of the inspection period, on October 30, 2018, the NRC staff responded (ML18260A203) to PSEGs LAR dated September 20, 2017, and issued License Amendment No. 213 that revised the technical specifications to adopt TSTF-542, Revision 2. Corrective Action Reference: 20792923 15 Enforcement: Violation: TS 3.6.5.1, Secondary Containment Integrity, requires that secondary containment integrity be maintained, and is applicable during OPDRVs. The required action for this specification without secondary containment integrity in this condition of applicability is to suspend OPDRVs. Contrary to the above, from April 19 through April 29, 2018, HCGS performed OPDRVs without secondary containment integrity. Therefore, set and maintain secondary containment integrity during OPDRVs without suspending the operation was considered a condition prohibited by TSs as defined by 10 CFR 50.73(a)(2)(i)(B). 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 and enforcement discretion was previously authorized by EA-2017-071; therefore, no enforcement action will be issued for this violation. The disposition of this violation closes LER 05000354/2018-001-00. |
| 05000331/FIN-2018002-02 | Duane Arnold | 2018Q2 | Minor Violation | Minor Violation: On June 19, 2016, while operating at 82 percent power, two secondary containment access airlock doors were opened simultaneously during surveillance testing as part of STP 3.6.4.102, Secondary Containment Airlock Verification. The inspectors determined this event was caused by inadequate procedural guidance which directed the user to attempt to open one airlock door while the other door was already open. During this test, the interlock failed because the permanent magnets had rotated and were misaligned. This failure could have been identified without challenging airlock interlock integrity if the second airlock door wasnt held open. The failure to have adequate procedural guidance for testing the secondary containment airlock doors was a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, which requires licensees to have procedures appropriate to the circumstance when performing safety-related activities. In response to this issue, the licensee immediately closed the airlock doors. In addition, the licensee submitted a TS change request to address the concurrent opening of two secondary containment airlock doors. The licensees corrective action program is tracking the TS change as CR 02034076, Secondary Containment Airlock Doors #225 and 228 Both Opened. Screening: The issue screened as minor because all of the questions associated with a minor issue found in IMC 0612, Appendix B were answered No due to the licensee reestablishing secondary containment operability immediately after the second airlock door opened. In addition, the inspectors considered the failure to have an appropriate procedure was less than a Severity Level IV violation in accordance with the NRCs Enforcement Policy. Violation: The failure to comply with 10 CFR Part 50, Appendix B, Criterion V, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes LER 05000331/2016001. |
| 05000293/FIN-2018002-05 | Pilgrim | 2018Q2 | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR 50.72(b)(3)(v)(C) requires licensees to a notify the NRC within 8 hours any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above, Entergy did not make a required notification pursuant to 10 CFR 50.72(b)(3)(v)(C). Specifically, on June 20, 2017, secondary containment was declared inoperable due to simultaneous opening of both airlock doors, and Entergy did not make the required notification until June 22, 2017. Significance/Severity: This violation is being treated under the NRCs traditional enforcement process, for impeding the regulatory process, specifically Entergy did not make a required notification, as outlined in Inspection Manual Chapter 0612, Appendix B. The Reactor Oversight Processs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The severity of this violation was determined to be Severity Level IV, as outlined in Example 9 from Section 6.9.d. of the NRC Enforcement Policy. Corrective Action References: CR-PNP-2017-06380 and CR-PNP-2017-07015 The disposition of this finding closes Licensee Event Report 2017-011-00. |
| 05000293/FIN-2018002-04 | Pilgrim | 2018Q2 | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions appropriate to the circumstances and shall be accomplished in accordance with the instructions. Contrary to the above, from January 1994 to June 2017, Entergy modified site surveillance procedure 8.M.3-18, Standby Gas Treatment System Exhaust Fan Logic Test and Instrument Calibration, without prescribing adequate documented instructions for the condition caused by the testing. Specifically, Entergy failed to identify that the procedurally prescribed lineup of the standby gas treatment system resulted in secondary containment being inoperable due to the large opening introduced into the system. Significance/Severity: The inspectors evaluated this finding using Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance. Corrective Action Reference: CR-PNP-2017-11714 The disposition of this violation closes Licensee Event Reports 05000293/2017-013-00 and 05000293/2017-013-01. |
| 05000387/FIN-2018002-04 | Susquehanna | 2018Q2 | 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. |
| 05000324/FIN-2018002-02 | Brunswick | 2018Q2 | Enforcement Action 18-080: Implementation of EGM 11-003, Revision 3, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements during Operations with a Potential for Draining the Reactor Vessel (OPDRV) | During the Unit 1 spring 2018 refueling outage, the OPDRVs activities are listed below: March 7, 2018: 148 gallons per minute (gpm) leakage associated with local leak rate testing (LLRT) of valves 1-G31-F001 and -F004.March 8, 2018: 82 gpm leakage for A RHR loop draining to support maintenance.March 12, 2018: 81.2 gpm leakage to replace the local power range monitors and intermediate power range dry tubes.March 14, 2018: 71.2 gpm leakage to replace the local power range monitors.March 15, 2018: 25 gpm leakage to replace A recirculation pump seal package.March 22, 2018: 25 gpm leakage to replace A recirculation pump seal package.March 27, 2018: 164 gpm leakage to facilitate control rod drive system venting.March 28, 2018: 288 gpm leakage to account for leakage past scram discharge and vent valves during testing.These activities took place without secondary containment being operable. Corrective Actions: EGM 11-003 allows enforcement discretion regarding secondary containment operability during Mode 5 OPDRV activities provided the licensee meets certain requirements. The licensee met the stipulations of the EGM by executing their procedure 1SP-16-100, EGM 11-003 OPDRV Activities, Rev 001, for each OPDRV activity during the Unit 1 Spring 2018 refueling outage. Additionally, as required by the EGM, the licensee submitted a license amendment request (BSEP 17-0060) on June 29, 2017. The amendment was approved on April 13, 2018, and will be implemented prior to the 2019 Unit 2 spring refueling outage. Corrective Action Reference: The issue was entered into the licensees corrective action program as NCR 2189536. Violation: TS 3.6.4.1, Secondary Containment, requires that secondary containment be operable and is applicable during OPDRVs. The required action if secondary containment is inoperable in this condition is to initiate actions to suspend OPDRVs immediately. Contrary to the above, on activities listed above, the licensee failed to maintain secondary containment operable while performing OPDRVs on Unit 1. Severity/Significance: According to EGM 11-003, the NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities provided the licensee meets certain requirements such as monitoring vessel level, maintaining capability to isolate leakage paths, providing minimum makeup flow rate, etc. These requirements provide a reasonable assurance of public health and safety during draining activities in Mode 5 while the secondary containment is inoperable 13 Enclosure Discretion Basis: The NRC exercised 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. These violations were identified during the discretion period described in EGM 11-003, Revision 3, and the licensee met the criteria established in the EGM prior to and during these activities. |
| 05000293/FIN-2018002-02 | Pilgrim | 2018Q2 | Loss of Secondary Containment Integrity due to Simultaneously Opened Airlock Doors | A self-revealed Green finding was identified when personnel did not implement a procedure requiring the closure and verification of doors credited with specific design functions. Procedure 1.3.135, Control of Doors, requires station personnel to ensure closing and latching of doors. Failure to meet this requirement caused the loss of secondary containment integrity and unplanned entry into Technical Specification (TS) condition 3.7.C.1. |
| 05000416/FIN-2018002-01 | Grand Gulf | 2018Q2 | Failure to Institute Effective Corrective Action to Preclude Repetition | An NRC-identified,Green non-cited violation of 10CFRPart50, AppendixB, CriterionXVI, Corrective Action, was identified when the licensee failed to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee left a secondary containment personnel hatch in an open configuration for approximately 30 minutes while performing a roof inspection, which rendered secondary containment inoperable. This issue had also previously occurred in 2016, but corrective actions to prevent it from occurring again were ineffective. |
| 05000327/FIN-2018001-03 | Sequoyah | 2018Q1 | 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: Sequoyah Unit 1 and Unit 2 Technical Specification 3.7.12, Auxiliary Building Gas Treatment System (ABGTS), requires two ABGTS trains be operable in modes 1, 2, 3, and 4. Contrary to the above, from March 3-7, 2017, the licensee blocked open door A212 resulting in the inoperability of the auxiliary building secondary containment enclosure boundary and thus inoperability of both trains of the ABGTS. Significance/Severity Level: Green. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building.Corrective Action Reference: CR1269767 |
| 05000373/FIN-2018001-03 | LaSalle | 2018Q1 | Enforcement Action (EA) 18035: Licensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance | On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015 (ML15111A269) and revised on February 7, 2017 (ML16355A286). The EGM applies specifically to an SSC that is determined to be inoperable for tornado generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. The EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within approximately 60 days of issue discovery. The comprehensive measures should remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Appendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants (GDC), Criterion 4, Environmental and Dynamic Effects Design Basis, states in part that SSCs important to safety shall be adequately protected against dynamic effects including missiles. On February 15, 2018, during evaluation of protection for Technical Specifications (TS) equipment from the damaging effects of tornado generated missiles, LaSalle County Station identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from tornado generated missiles. Specifically, tornado generated missiles could strike the components supporting the operation of Control Room (VC) and Auxiliary Electric Room (VE) ventilation. This could result in inoperable VC/VE systems, which provide a protected environment for occupants to control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke if a tornado were to occur. In addition, the Unit 2 Division 2 motor control center (MCC) 236X1 was affected, which impacted various loads on Unit 2 including the Unit 2 standby gas treatment, Unit 2 Division 2 post LOCA system, B main control room area filtration system supply and exhaust fan, reactor building Division 2 isolation damper control logic, Unit 2 Division 2 battery room exhaust fan and Unit 2 24/48 Volt battery rooms exhaust fans. This would result in a loss of power to components and systems rendering them inoperable. The condition was reported to the NRC in Event Notice (EN) 53213 as an unanalyzed condition and potential loss of safety function. Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS Limiting Conditions for Operation (LCOs) in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of the implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. Initial (immediate) compensatory measures were established by an operations standing order that included: Procedures were verified to be put in place, with associated current training, for performing actions in response to a tornado. Procedures were verified to be put in place, with associated current training, for actions to be taken if a tornado watch is issued for the area. Procedures were verified to be put in place, with associated current training, for actions to be taken if a tornado warning is issued for the area. Verification that training was up to date for individuals responsible for implementing preparation and response procedures; and Established a heightened station awareness and preparedness level relative to identified tornado missile vulnerabilities. The comprehensive (60 day) compensatory measures were established by incorporating the standing order actions and adding additional detail to operating procedure LOATORN001, High Winds/Tornado, Revision 22, for completing additional inspections and restoration actions on equipment vulnerable to tornado missile damage. Corrective Action Program References: AR 4104401; AR 4104391; AR 4104393; AR 4104396; AR 4104397. Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.7.4, Control Room Area Filtration (CRAF) System; TS 3.7.5, Control Room Area Ventilation Air Conditioning (AC); TS 3.6.4.2, Secondary Containment Isolation Valves (SCIVs); TS 3.6.4.3; Standby Gas Treatment (SGT) System; and TS 3.8.7, Distribution SystemsOperating. Severity/Significance: The subject of this enforcement discretion, associated with tornado missile protection deficiencies was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS, Accession No. ML16355A286. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions were to remain in place until permanent repairs were completed, which for LaSalle were required to be completed by June 10, 2018, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed |
| 05000397/FIN-2018001-01 | Columbia | 2018Q1 | Failure to Follow Procedure Leads to Loss of Secondary Containment | The inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to perform maintenance in accordance with documented instructions appropriate to the circumstances. Specifically, on September 12, 2017, the failure to verify power sources per Work Order 02072924 caused an electrical transient that caused the reactor building exhaust valve and supply valve to lose power and close, resulting in a loss of secondary containment |
| 05000315/FIN-2017004-02 | Cook | 2017Q4 | Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown | Refueling Outage Activities a. Inspection Scope The inspectors reviewed the Outage Safety Plan (OSP) and contingency plans for the Unit 1 refueling outage (RFO), conducted September 13 through November 26, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. During the RFO, the inspectors observed portions of the shutdown and cooldown processes and monitored licensee controls over the outage activities listed below: licensee configuration management, including maintenance of defense-in-depth commensurate with the OSP for key safety functions and compliance with the applicable TS when taking equipment out of service; implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing; installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error; controls over the status and configuration of electrical systems to ensure that TS and OSP requirements were met, and controls over switchyard activities; monitoring of decay heat removal processes, systems, and components; controls to ensure that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system; reactor water inventory controls including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss; controls over activities that could affect reactivity; maintenance of secondary containment as required by TS; licensee fatigue management, as required by 10 CFR 26, Subpart I; refueling activities, including fuel handling and reactor assembly/disassembly; startup and ascension to full power operation, tracking of startup prerequisites, walkdown of the containment to verify that debris had not been left which could block emergency core cooling system suction strainers, and reactor physics testing; and licensee identification and resolution of problems related to RFO activities. Documents reviewed are listed in the Attachment to this report. Inspections activities performed in the third quarter coupled with those in the fourth quarter constituted one RFO sample as defined in IP 71111.2005. b. Findings (Opened) Unresolved Item 05000315/201700402, Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown Introduction: Shortly after the shutdown for the Unit 1 refueling outage in September 2017, the licensee was establishing conditions in the charging and letdown system for the upcoming cooldown. After lowering letdown flow and attempting to adjust pressure, a letdown safety valve lifted and failed to completely reseat. Review of plant parameters following the event revealed that the evolution created saturation conditions in the letdown system. Subsequently, the steam bubbles collapsed causing a water hammer that lifted and damaged a relief in the system. The event was discussed in Section 4OA3 of Inspection Report 05000315/05000316/2017003. Description: The inspectors reviewed the licensees follow up of the issue in the CAP and spoke to personnel in the operations and maintenance departments. The licensee identified potential issues in the areas of procedure adequacy, operator performance, and equipment performance. However, the inspectors could not reconcile information on plant conditions with licensees statements regarding the cause. Because of ambiguity regarding the cause, the inspectors could not determine whether the corrective actions taken by the licensee were adequate. The licensee determined that an apparent cause evaluation need not be done therefore the inspectors reviewed available data, including plant computer data and a prior event from 2004. Since it is unclear what, if any, performance deficiency exists associated with this issue, the inspectors determined an unresolved item (URI) was necessary pending further follow up of the issue.Following the lifting of the safety valve, the licensee isolated letdown to stop the remaining leakage through the valve. The licensee then cycled the valve sufficiently enough for it to reseat so letdown could be restored and the cooldown continued. The safety valve was later discovered to be damaged from the event, so it was also repaired. Walkdowns were also conducted of the letdown piping to ensure no damage had occurred during the pressure transient. As part of their corrective actions, the licensee made some changes to the letdown procedure, recalibrated a letdown flow control valve, and developed actions to cover the event and lessons-learned in training. However, as stated above, the inspectors were unable to determine if these were sufficient to address the prevailing cause of the issue. The inspectors developed a series of questions for the licensee to explore more of the details behind the various potential issues. In order close the URI, the inspectors need to review the licensees response to questions provided and review available documentation of the event. (URI 05000315/201700402, Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown) |
| 05000461/FIN-2017003-05 | Clinton | 2017Q3 | Failure to Establish Secondary Containment Prior to Entering MODE 2 | The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS LCO 3.0.4, for the failure to follow station procedure CCAA201, Plant Barrier Control Program, Revision 11. Specifically, the licensee entered MODE 2 from MODE 4 without meeting the requirements of LCO 3.0.4 for entering a mode when an applicable LCO is not met. The licensee had not met LCO 3.6.4.1 because the doors to the B reactor water cleanup room were both opened instead of being closed to make secondary containment operable as required in MODE 2. The licensee entered this issue into their CAP as AR 04017613. As corrective actions, the licensee planned to conduct training for site personnel.The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to follow the station procedure by not identifying that the open doors required a plant barrier impairment (PBI) permit that would have identified the doors as a constraint to entering MODE 2 resulted in the unit transitioning to MODE 2 with the secondary containment inoperable. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone and determined 5 to be of very low safety significance because the finding only represented a degradation of a radiological barrier function provided for auxiliary building. The inspectors determined that this finding affected the cross-cutting are of human performance in the aspect of training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent work force and instill nuclear safety values. Specifically, station personnel did not know the process for routing a PBI permit and did not know when a PBI permit was required. (H.9) |
| 05000354/FIN-2017007-01 | Hope Creek | 2017Q3 | Secondary Containment Integrity Not Maintained Due to Door not Properly Dogged | Green. The team identified a Green, non-cited violation (NCV) of Technical Specification (TS) 3.6.5.1, for failure to maintain secondary containment integrity. Specifically, while Hope Creek station was operating in mode 1, PSEG personnel did not ensure secondary containment door R-4302 was properly latched (dogged) closed in accordance with procedure CC-AA-201, Plant Barrier Control Program . The licensees failure to ensure the door was properly dogged closed was a performance deficiency and resulted in a degraded secondary containment barrier for approximately 44 hours. The team determined that PSEG operated in violation of the TS LCO which requires restoration of secondary containment integrity within 4 hours or be in at least hot shutdown within the next 12 hours and in cold shutdown within the following 24 hours. Following identification of the door condition by the team PSEG personnel properly dogged the door closed restoring secondary containment. This finding was determined to be of more than minor significance because it is associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, in its un-dogged position the door would not have remained closed, as required to maintain secondary containment integrity, during all design basis accidents. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power , Exhibit 3, Barrier Integrity Screening Questions, the team determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of primary reactor containment (valves, airlocks , etc.), containment isolation system (logic and instrumentation), and heat removal components. The finding was determined to be associated with the cross-cutting area of Human Performance - Procedure Adherence (H.8), in that, licensee personnel did not follow process, procedures, and work instructions which required the secondary containment door to be closed and dogged. |
| 05000321/FIN-2017002-04 | Hatch | 2017Q2 | Licensee-Identified Violation | TS 3.6.4.1 requires secondary containment be operable in Mode 1 and during movements of irradiated fuel assemblies in the secondary containment. Contrary to the above, on February 8 at 1035, with Unit 1 operating at 100 percent RTP and Unit 2 conducting refueling operations, secondary containment was made inoperable when Unit 2 reactor building containment was breached for a scheduled refueling outage and a configuration control error on the Unit 2 standby gas treatment system provided a uncontrolled opening into the secondary containment for the Unit 1 reactor building and the common refueling floor. A temporary blind flange had been incorrectly installed on the upstream side vice downstream side of the Unit 2 standby gas treatment inlet isolation valve when the valve had been removed from the system for testing. This configuration rendered secondary containment for the Unit 1 reactor building and the common refueling floor inoperable. A senior reactor operator performing a plant tour noted the incorrect flange configuration and at 2017 on February 17, the blind flange was moved to the downstream side of the Unit 2 standby gas treatment inlet isolation valve to restore compliance. Inspectors screened the finding in accordance with IMC 609 Appendix A The Significance Determination Process (SDP) for Findings at-Power. The finding screened as very low safety significance (Green) because the questions in Appendix A Exhibit 3 for Control Room, Auxiliary, Reactor, or Spent Fuel Pool Building, were answered no. This issue was documented in the licensees corrective action program as CR 10332592. |
| 05000293/FIN-2017002-07 | Pilgrim | 2017Q2 | Untimely 10 CFR 50.72 Notification of a Secondary Containment System Functional Failure | An NRC-identified SL IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, was identified because both trains of the SBGTS were made inoperable during surveillance testing, and the condition was not reported to the NRC within eight hours of the occurrence, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. Specifically, on April 5, 2017, while performing TS SR 4.7.C, trains A and B of the SBGTS were made inoperable leading to the inoperability of the Secondary Containment System (SCS). As a corrective action, Entergy personnel performed a causal evaluation. This issue was entered into the CAP as CR 2017-7446. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a condition that could have prevented a safety function was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy (the failure of a licensee to make a report as required by 10 CFR 50.72 or 10 CFR 50.73), the inspectors determined that the violation was a SL IV violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, inspectors did not assign a cross-cutting aspect, in accordance with IMC 0612, Appendix B. |
| 05000293/FIN-2017002-05 | Pilgrim | 2017Q2 | Damper Failure Causes Loss of Secondary Containment | A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and TS 3.7.C.2, Containment Systems Secondary Containment, was identified because Entergy did not establish an appropriate interval to overhaul the secondary containment isolation dampers. As a result, the refueling floor supply isolation dampers were operated beyond the recommended overhaul interval and subsequently failed. Entergys corrective actions included cleaning, lubricating, and post-work testing the failed refueling floor supply isolation dampers. This issue was entered into the CAP as CR 2017-0494. The performance deficiency is more than minor because it is associated with the 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, Entergys preventative maintenance (PM) for the refueling floor supply isolation dampers was inadequate to ensure the availability and reliability of SSCs required to maintain secondary containment operable. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency only represented a degradation of the radiological barrier function provided by the reactor building and standby gas treatment system (SBGTS). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution - Resolution, in that Entergy personnel did not take effective corrective actions to address issues in a timely manner. Specifically, in 2016, Entergy personnel identified there were deficiencies in the PM program with technical justifications for deferring PMs. Entergy reasonably had the opportunity to identify which PMs were not performed within recommended guidelines and make appropriate changes as needed. (P.3) |
| 05000293/FIN-2017002-06 | Pilgrim | 2017Q2 | Secondary Containment Testing not performed per Technical Specifications | An NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, and TS 4.7.C, Containment Systems Secondary Containment, was identified when Entergy performed a surveillance test requiring a refueling outage while online. Specifically, Entergy performed Procedure 8.7.3, Secondary Containment Leak Rate Test, TS Surveillance Requirement (SR) 4.7.C from February 27, 1997, to April 5, 2017. As corrective actions, Entergy re-performed the test during the April 2017 refueling outage prior to refueling. This issue was entered into the CAP as CR 2017-2900. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protects the public from radionuclide releases caused by accidents or events. Specifically, Entergy intentionally removed the safety function of standby gas and secondary containment for operational convenience and did not comply with the requirements of TS SR 4.7.C which requires the test to be performed during a refueling outage before refueling. In accordance with IMC 0609.04, Initial Characterization of Findings, issued October 7, 2016, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green), because the finding only represented a degradation of the radiological barrier function provided for the SBGTS. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance - Conservative Bias, in that Entergy personnel did not use decision making-practices that emphasize prudent choices over those that are simply allowable. Specifically, operators did not refer to the TSs to understand the required conditions for a secondary containment surveillance test. Operators followed an inadequate site procedure for the plant conditions at the time and did not question why removal of a safety function for operational convenience was acceptable. (H.14) |
| 05000366/FIN-2017002-02 | Hatch | 2017Q2 | Performance of Operations with Potential to Drain the Reactor Vessel (OPDRV) Without Secondary Containment | The inspectors reviewed this LER for potential performance deficiencies and/or violations of regulatory requirements. In February 2017, during the Unit 2 refueling outage, operations with the potential to drain the reactor vessel (OPDRV) activities were performed while in Mode 5 (Refueling Mode) contrary to Technical Specification (TS) 3.6.4.1. Enforcement Guidance Memorandum (EGM) 11-003, Revision 3, provided required interim actions which were incorporated into procedure 31GO-OPS-025-0 Operations with the Potential to Drain the Reactor Vessel. This procedure was used during the OPDRV activities for the Unit 2 refueling outage. LER 05000366/2017-001- 00 is closed. Description: The inspectors reviewed the plants implementation of Enforcement Guidance Memorandum 11-003 during maintenance activities which had the potential to drain the reactor vessel during the Unit 2 refueling outage. The activities were: Local power range monitors removal and replacement February 10, 2017; Control rod drive insert / recouple activity February 11, 2017; and Hydraulic Control Unit Venting February 12-13, 2017. 15 These activities took place without secondary containment being operable. Inspectors verified compliance with the guidelines of Enforcement Guidance Memorandum 11-003 prior to and during these activities. This condition was documented in the licensees corrective action program as CR 10329405, 10329857, 10330152, and 10330153. Enforcement: Unit 2 TS 3.6.4.1 required, in part, that activities that had the potential to drain the reactor vessel be conducted only with secondary containment operable. Contrary to that requirement, the licensee conducted activities that could cause the reactor vessel to drain while secondary containment was inoperable. The NRC is exercising enforcement discretion (Enforcement Action (EA)-17-124) in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy because the violation was identified during the discretion period described in Enforcement Guidance Memorandum 11-003. Therefore, the NRC will not issue enforcement action for this violation, subject to the license amendment request which was submitted on April 20, 2017. |
| 05000440/FIN-2017002-02 | Perry | 2017Q2 | 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. |
| 05000324/FIN-2017002-01 | Brunswick | 2017Q2 | Implementation of EGM 11- 003, Revision 3, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential for Draining the Reactor Vessel | Implementation of EGM 11- 003, Revision 3, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential for Draining the Reactor Vessel a. Inspection Scope The inspectors reviewed the plants implementation of NRC EGM 11 -003, Revision 3, during Unit 2 maintenance activities for operations with a potential for draining the reactor vessel (OPDRVs) , during the Unit 2 refueling outage. Inspectors verified that for all dates, all other TS requirements were met during OPDRVs with secondary containment inoperable. Documents reviewed are listed in the Attachment. b. Findings Description . During the Unit 2 refueling outage, the OPDRVs activities are listed below: March 21, 2017: 6 gallons per minute leakage for maintenance associated with the 2B recirculation pump seal rebuild March 23, 2017: 48 gallons per minute leakage for hydraulic control unit draining to support rod maintenance March 24, 2017: 65 gallons per minute leakage to close the excess flow check valve and cap the drain line March 27, 2017: 71.2 gallons per minute leakage to replace the low power range monitors and intermediate power range monitors These activities took place without secondary containment being operable. Enforcement . TS 3.6.4.1, Secondary Containment, requires that secondary containment be operable and is applicable during OPDRVs. The required action if secondary containment is inoperable in this condition i s to initiate actions to suspend OPDRVs immediately. Contrary to the above, on March 21, 2017, March 23, 2017, March 24, 2017 , and March 27, 2017 , the licensee failed to maintain secondary containment operable while performing OPDRVs. However, because the violations were identified during the discretion period described in EGM 11 -003, Revision 3, and the licensee met the criteria established in the EGM prior to and during these activities, the NRC exercised enforcement discretion (Enforcement Action -17- 123 ) for the dates of March 21, 2017, March 23, 2017, March 24, 2017, and March 27, 2017, in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforc ement action for this v iolation. The inspectors observed that Brunswick has already submitted a license amendment request (BSEP 17- 0060) on June 29, 2017 which was accepted for review by the NRC on July 18, 2017. The licensee entered this issue into the CAP as NCR 2110409. |
| 05000387/FIN-2017001-01 | Susquehanna | 2017Q1 | 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) |
| 05000237/FIN-2017001-02 | Dresden | 2017Q1 | Secondary Containment Inoperability Due to Lapse in Procedure Use and Adherence | A self-revealed finding of very low safety significance (Green) and associated NCV of Technical Specification (TS) 5.4.1, Procedures, occurred on November 8, 2016, due to the licensees failure to follow procedures designed to ensure secondary containment integrity, when reactor building (RB) pressure relative to the outside environment was less than 0.25 inches water column (in WC) vacuum as required by TS 3.6.4.1, Secondary Containment. Specifically, work group personnel did not communicate to operations regarding degraded sealing surfaces on the RB Equipment Access outer door as required by procedure DAP 1303, Unit 2 Reactor Building Trackway Interlock Door Access Control, therefore when standby gas treatment (SBGT) started as a part of a planned surveillance test, vacuum lowered, rendering secondary containment inoperable. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Barrier Integrity Cornerstone Attribute of Human Performance and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the drop in secondary containment differential pressure to less than 0.25 in WC vacuum, resulted in a loss of secondary containment and failure of its safety function as specified by TS 3.6.4.1 and Updated Final Safety Analysis Report (UFSAR) section 6.2.3. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. The inspectors reviewed the Barrier Integrity Screening Questions in Appendix A, Exhibit 3 and answered Yes to question C.1. As a result, the finding was determined to have very low safety significance (Green). This finding has a cross cutting aspect in the area of Problem Identification and Resolution, Identification, because individuals failed to identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee did not report a condition adverse to quality with regards to degraded seals on the RB equipment access outer door to operations as required by procedure DAP 1303, therefore not ensuring secondary containment integrity. |
| 05000354/FIN-2017001-03 | Hope Creek | 2017Q1 | Operations with a Potential to Drain the Reactor Vessel (OPDRV) Without Secondary Containment | On October 23, 24, and 31, 2016, during a planned refueling outage with the reactor cavity flooded up in Mode 5, Hope Creek conducted multiple OPDRVs without an operable secondary containment. The conduct of an OPDRV without establishing secondary containment integrity is a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B). Secondary containment is required by TS 3/4.6.5.1 in Operational Condition (*), which is a condition when recently irradiated fuel is being handled during an OPDRV. The required action for this specification is to suspend handling recently irradiated fuel and OPDRV operations. In this case, the specific OPDRVs were control rod drive mechanism replacements (8:40 a.m. on October 23, 2016, through 10:50 p.m. on October 23, 2016), local power range monitor replacements (10:50 p.m. on October 23, 2016, through 8:07 a.m. on October 24, 2016), additional control rod drive mechanism and local power range monitor replacements (8:07 a.m. on October 24, 2016, through 8:23 a.m. on October 25, 2016), and the fill and vent for the A and B RRP seal (11:21 a.m. on October 31, 2016, through 12:02 p.m. on November 1, 2016). The OPDRVs were completed in accordance with PSEG procedure OP-HC-108-102, "Management of Operations with the Potential to Drain the Reactor Vessel (OPDRV)," Revision 5, dated October 6, 2016. These OPDRVs were completed and exited at 12:02 p.m. on November 1, 2016. The NRC issued EGM 11-003, Revision 3, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential For Draining the Reactor Vessel, on January 15, 2016, which provides, in part, for the exercise of enforcement discretion only if the licensee demonstrates that it has implemented specific interim actions during any OPDRV activity. The inspectors determined that PSEGs implementation of these specific interim actions during these OPDRV activities were adequate and met the intent of EGM 11-003, Revision 3. The inspectors assessments of PSEGs implementation of these criteria during each of the multiple OPDRV activities are described below: The inspectors observed that, as required by the EGM, the OPDRV activity was logged in the control room narrative logs and that the log entry appropriately recorded the safety-related pump (B RHR) that was the standby source of makeup designated for the evolution. The inspectors noted that the reactor vessel water level was maintained at least 22 feet and 2 inches over the top of the reactor pressure vessel (RPV) flange in compliance with the minimum water level allowed by Hope Creek TS limiting condition for operation (LCO) 3.9.8 applicability. The inspectors also noted that at least one safety-related pump was the standby source of makeup designated in the control room narrative logs for the evolution with the capability to inject water equal to, or greater than, the maximum potential leakage rate from the RPV for a minimum time period of 4 hours. PSEG reported that the worst case estimated time to drain the reactor cavity to the RPV flange was 36.6 hours, which met the EGM criteria of greater than 24 hours. The inspectors verified that the OPDRV was not conducted in Mode 4 and that PSEG did not move recently irradiated fuel during the OPDRV. The inspectors noted that PSEG had in place a contingency plan for isolating the potential leakage path. The inspectors verified that two independent means of measuring RPV water level (one alarming) were available for identifying the onset of loss of inventory events with sufficient time to close secondary containment before reactor water level reached the top of the RPV flange. Technical Specification 3.6.5.1 is applicable in Operational Conditions 1, 2, 3 and (*). This TS requires that secondary containment integrity shall be maintained. Operational Condition (*) is defined, in part, as being during OPDRV. TS 3.6.5.1, action b, states, in part, in operational condition (*) suspend operations with a potential for draining the reactor vessel. Contrary to the above, between 8:40 a.m. on October 23, 2016, and 12:02 p.m. on November 1, 2016, Hope Creek Generating Station did not maintain secondary containment integrity while conducting OPDRV activities. Because the violation was identified during the discretion period described in EGM 11-003 Revision 3, the NRC is exercising enforcement discretion in accordance with NRC Enforcement Policy Section 2.2.4, Exceptions to Using Only the Operating Reactor Assessment Program, and Section 3.5, Violations Involving Special Circumstances, and, therefore, will not issue enforcement action for this violation. In accordance with EGM 11-003 Revision 3, each licensee that receives discretion must submit a license amendment request within 4 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the STS to provide more clarity to the term OPDRV. The inspectors observed that PSEG is tracking the need to submit a license amendment request in its CAP as NOTF 20559547 (Order 70138857). No findings were identified. This LER is closed. |
| 05000388/FIN-2016004-03 | Susquehanna | 2016Q4 | 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) |
| 05000461/FIN-2016009-02 | Clinton | 2016Q4 | Failure to Scope SFP Temperature and Level Instruments into the Maintenance Rule Program | The team identified a finding of very-low safety significance (Green) and an associated NCV of Paragraph (b)(2)(i) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensee failure to scope non-safety related mitigating structure, systems, and components (SSCs) used within an emergency operating procedure (EOP) into Maintenance Rule Program. Specifically, an EOP used spent fuel pool (SFP) low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. The licensee captured the team concerns in their CAP as AR 02736193, performed an extent of condition to identify any other SSC addition to the EOPs requiring them to be added to the Maintenance Rule Program scope, and initiated plans to incorporate the affected SSCs into the Maintenance Rule Program scope. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of SSC performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that SSCs within its scope perform reliably and are capable of providing their intended Maintenance Rule function(s). In the case of the SFP temperature instruments, the licensee was not performing preventive maintenance to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm EOP entry conditions such that mitigating actions could be implemented to preserve secondary containment. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not use a systematic process for evaluating and implementing changes when updating the affected EOP in 2015. |
| 05000341/FIN-2016004-04 | Fermi | 2016Q4 | Inadequate Testing of SGTS Filters | The inspectors identified a finding of very low safety significance with an associated NCV of TS 5.5.7, Ventilation Filter Testing Program. The licensee failed to perform testing of the standby gas treatment system (SGTS) high-efficiency particulate air (HEPA) filters that demonstrated a penetration and system bypass of less than 0.05 percent. The licensee entered this violation into its CAP as CARD 1628812. The licensee declared the Division 1 SGTS subsystem inoperable until testing was performed satisfactorily and evaluated the extent of condition on the control room filtration system. This performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute for the control room and auxiliary building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not adequately testing the SGTS HEPA filters, the ability of the SGTS to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured. The finding was determined to be of very low safety significance because it involved only a degradation of the radiological barrier function provided by the SGTS. The inspectors concluded that because this condition has existed for greater than three years, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified. |
| 05000341/FIN-2016004-06 | Fermi | 2016Q4 | Licensee-Identified Violation | Title 10 of the CFR, Section 50.72(a)(1)(ii), requires, in part, that the licensee shall notify the NRC Operations Center via the Emergency Notification System of those non-emergency events specified in Paragraph (b) that occurred within three years of the date of discovery. 10 CFR 50.72(b)(3) requires, in part, that the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the applicable conditions. 10 CFR 50.72(b)(3)(v)(C) requires, in part, that the licensee report any event or condition, that at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. 10 CFR 50.73(a)(1) requires, in part, that the licensee submit an LER for any event of the type described in this paragraph within 60 days after the discovery of the event. 10 CFR 50.73(a)(2)(v)(C) requires, in part, that the licensee report any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above: 1. Between September 1, 2013 and September 30, 2016, the licensee failed to notify the NRC Operations Center via the Emergency Notification System of numerous non-emergency events specified in Paragraph (b) within eight hours of the events. These events involved the loss of safety function of the secondary containment when secondary containment pressure exceeded the TS limit due to known effects of high winds. 2. The licensee failed to submit required LERs within 60 days after the discovery of numerous events between September 1, 2013 and September 30, 2016. These events involved the loss of safety function of the secondary containment when secondary containment pressure exceeded the TS limit due to known effects of high winds. Violations of 10 CFR 50.72 and 10 CFR 50.73 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to make reports to the NRC as required by 10 CFR 50.72(a)(1)(ii) and 10 CFR 50.73(a)(1). The licensee entered this violation into its CAP as CARD 16-27023. Title 10 of the CFR, Section 20.1501, requires, in part, that each licensee shall make, or cause to be made, surveys of areas that may be necessary for the licensee to comply with the regulations in this part and are reasonable for the circumstances to evaluate the magnitude and extend of radiation levels and the potential radiological hazards of the radiation levels. 10 CFR 20.1902(b) states that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words CAUTION, HIGH RADIATION AREA or DANGER, HIGH RADIATION AREA. Contrary to the above, on August 17, 2016, the licensee failed to conduct reasonable surveys to evaluate radiation levels to ensure compliance with the posting requirements of 10 CFR 20.1902(b) during activities known to cause changes in radiation levels. Specifically, the licensee failed to ensure surveys were performed while draining the annulus of the multi-purpose canister, which is an evolution known to change radiological conditions. An unposted high radiation area was identified several hours later when radiation protection personnel entered the area to perform surveys to ensure compliance with the containers Certificate of Compliance. This violation was entered into the licensees CAP as CARD 16-26586. The finding was assessed in accordance with IMC 0609, Appendix C, Occupational Radiation Safety SDP and determined to be of very-low safety significance because it did not involve as-low-as-reasonably-achievable planning or work controls, there was no overexposure nor substantial potential for an overexposure, and the ability to assess dose was not compromised |
| 05000298/FIN-2016004-03 | Cooper | 2016Q4 | Failure to Maintain Service Water Pump Maintenance Procedure | The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 3.6.4.2, Secondary Containment Isolation Valves, for the licensees failure to maintain secondary containment isolation valve HV-AOV-265 operable as a result of erecting scaffolding that interfered with valve operation. Specifically, between June 29, 2016, and September 14, 2016, the licensee erected scaffolding in close proximity of valve HV-AOV-265, such that, during valve stroking, the scaffolding would pinch the actuator air line and prevent the valve from closing, rendering the valve inoperable for approximately 10 weeks. This resulted in the licensees need to reduce power to approximately 50 percent in order to comply with technical specifications upon discovery. Immediate corrective actions included removal of the scaffolding, replacement of the pinched air line, and restoration of the valve to operable status. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-05608 and initiated a root cause evaluation to investigate this condition. The licensees failure to implement Procedure 7.0.7, Scaffolding Construction and Control, Revision 34, to ensure scaffolding did not adversely affect plant equipment, in violation of Technical Specification 3.6.4.2, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the structure, system, and component and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the improperly erected scaffolding prevented the operation of a secondary containment isolation valve, rendering it inoperable for approximately 10 weeks. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room, reactor building, spent fuel pool building, or standby gas treatment system. The finding had a cross-cutting aspect in the area of human performance associated with resources. Specifically, the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety (H.1). |
| 05000336/FIN-2016002-01 | Millstone | 2016Q2 | Secondary Containment Inoperability Due to Inadequate Procedures | The inspectors documented a self-revealing Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Dominion did not develop a Unit 3 supplementary leak collection and release system (SLCRS) damper procedure that was adequate to prevent the inoperability of the system. Specifically, deficiencies in procedure SP 3614I.3A, Supplementary Leak Collection and Release System Boundary Isolation Damper Test, as well as the SLCRS damper monitoring program and preventative maintenance strategy, led to both trains of the Unit 3 SLCRS failing their respective surveillance tests resulting in the inoperability of secondary containment. After the issue was identified, Dominion entered the condition into their corrective action program (CAP) as condition report (CR)1033408, declared the secondary containment inoperable until the plant entered a mode of technical specifications non-applicability, and conducted walkdowns and repairs to the system to restore it to compliance. This performance deficiency was considered to be more than minor because it adversely affected the system, structure, and component (SSC) and barrier performance attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, inadequate maintenance of the SLCRS system led to a system differential pressure during operation that was not adequate to meet its design basis surveillance requirement and thus rendered the system inoperable. Additionally, the performance deficiency was similar to IMC 0612, Appendix E, minor example 2.a. The finding was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined to be of very low safety significance (Green) since it only represented a degradation of the radiological barrier function provided for the auxiliary building. The finding is related to the cross-cutting aspect of Human Performance, Design Margins, because Dominion did not operate and maintain equipment within design margins. Specifically, Dominion did not appropriately monitor and maintain the SLCRS system in such a way that declining damper performance trends were identified and prevented prior to the inoperability of the system. |
| 05000321/FIN-2016002-03 | Hatch | 2016Q2 | Performance of Operations with Potential to Drain the Reactor Vessel (OPDRV) Without Secondary Containment | In February 2016, during the Unit 1 refueling outage, operations with the potential to drain the reactor vessel (OPDRV) activities were performed while in Mode 5 (Refueling Mode) contrary to Technical Specification (TS) 3.6.4.1. These OPDRV activities were also performed during the Unit 2 Refueling Outage. Enforcement Guidance Memorandum (EGM) 11-003, Revision 3, provided required interim actions which were incorporated into procedure 31GO-OPS-025-0 Operations with the Potential to Drain the Reactor Vessel. This procedure was used during the OPDRV activities for the Unit 1 refueling outage. Enforcement: Unit 1 TS 3.6.4.1 required, in part, that activities that had the potential to drain the reactor vessel be conducted only with secondary containment operable. Contrary to that requirement, the licensee conducted activities that could cause the reactor vessel to drain while secondary was inoperable. The inspectors determined this was a Severity Level IV violation. The NRC is exercising enforcement discretion (Enforcement Action (EA)-16-158) in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy because the violation was identified during the discretion period described in Enforcement Guidance Memorandum 11-003. Therefore, the NRC will not issue enforcement action for this violation, subject to a timely license amendment request. |
| 05000416/FIN-2016002-01 | Grand Gulf | 2016Q2 | Failure to Maintain Secondary Containment Operable during Roof Inspections | The inspectors identified a Green, non-cited violation of Technical Specification Surveillance Requirement 3.0.1, for the failure to meet Surveillance Requirement 3.6.4.1.1 and declare Limiting Condition for Operation 3.6.4.1 not met. Specifically, the licensee did not maintain the enclosure building hatch penetration in the closed position as required by Surveillance Requirement 3.6.4.1.1, which resulted in secondary containment being inoperable. The licensee restored compliance by closing the hatch following the surveillance, and put corrective actions in place to control the enclosure building hatch penetration in a closed position except for entry and exit for the inspection. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-1-2016-03707. The failure to declare that Limiting Condition for Operation 3.6.4.1 was not met when the enclosure building hatch was maintained in the open position was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, on April 7, 2016, the licensee did not maintain the enclosure building hatch penetration in the closed position as required by SR 3.6.4.1.1, which resulted in secondary containment being inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool, or standby gas treatment (SBGT) system (BWR). This finding has a cross-cutting aspect in the area of human performance associated with documentation, in that, the organization failed to create and maintain complete, accurate and up-to-date documentation. Specifically, Work Order 52671695 for implementing the roof inspection was not complete and accurate with regards to the impact on operability of secondary containment when leaving the enclosure building hatch penetration open during inspection activities. |
| 05000352/FIN-2016001-01 | Limerick | 2016Q1 | Reactor Enclosure Recirculation System Design Change not Evaluated | A self-revealing Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50 (10 CFR 50), Appendix B, Criterion III, Design Control, was identified because Exelon did not properly maintain the design of the LGS Unit 1 reactor enclosure recirculation system (RERS). Specifically, Exelon replaced the Unit 1 1A RERS flow straightener assembly using thinner material than was originally qualified and did not evaluate the change in design. Exelon initiated IR 2563872 and implemented a temporary configuration change that removed the flow straightener assembly from the system and restored Unit 1 RERS to operability on October 5, 2015. Exelon also initiated corrective actions to install a new flow straightener assembly with correctly sized honeycomb material. This finding is more than minor because it adversely affected the design control attribute of the barrier integrity cornerstone to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the inadequate 1A RERS flow straightener assembly installed in 2012 resulted in degraded performance and then unplanned unavailability of 1A RERS from October 1 to 5, 2015. Using IMC 0609, Appendix A, Exhibit 3, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the degraded 1A RERS performance and associated unavailability only represented a degradation of the radiological barrier function provided for the standby gas treatment system and screened to Green. The inspectors determined that the finding did not have cross-cutting aspect because the performance deficiency did not occur within the last three years, and the inspectors did not conclude that the primary cause of the performance deficiency represented present Exelon performance. |
| 05000410/FIN-2016001-05 | Nine Mile Point | 2016Q1 | Licensee-Identified Violation | The holder of an operating license under this part shall submit a Licensee Event Report (LER) for any event of the type described in this paragraph within 60 days after the discovery of the event. (v) Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material. Contrary to the above from June 2, 2014, until October 5, 2015, Exelon failed to submit an LER notification to the NRC within 60 days after discovery of a condition which could have prevented the safety function of a SSC needed to control the release of radioactivity on April 2, 2014 at 11:20 a.m. Specifically, secondary containment being declared inoperable due to both airlock doors being open at the same time in Mode 5 with an OPDRV in progress. The inspectors reviewed the violation using IMC 0612, Appendix B and the NRC Enforcement Policy. This violation impacted the regulatory process so traditional enforcement applies. Comparing this violation to the examples in the NRC Enforcement Policy Chapter 6, the violation matches Severity Level IV Example 6.9.d.9, a licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. The NRC did not rely upon the information to make any regulatory decisions, and the error did not result in increased scope or effort of NRC inspections. Compliance was restored when Exelon submitted LER 05000410/2014-007-01 to correct the public record and inform the NRC. Exelon staff entered the issue into its CAP. |
| 05000333/FIN-2016001-04 | FitzPatrick | 2016Q1 | Untimely 10 CFR 50.72 Notification of Inoperable Secondary Containment | The inspectors identified a SL IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because unplanned inoperability of the secondary containment system was not reported to the NRC within eight hours of the occurrence, as required by 10 CFR 50.72(b)(3)(v), Event or Condition That Could Have Prevented Fulfillment of a Safety Function. Specifically, following reasonable resolution of questions regarding the reliability of secondary containment differential pressure (d/p) instrumentation indications, FitzPatrick staff did not promptly report that, during a transfer from normal reactor building ventilation in service to the reactor building being isolated with the SBGTS in service, reactor building d/p briefly dropped below the TS required minimum value of 0.25 inches of vacuum water gauge and therefore caused the secondary containment system to be inoperable. As immediate corrective action, the event was reported to the NRC in accordance with 10 CFR 50.72(b)(3)(v). The issue was entered into the CAP as CR-JAF-2015-05244 and CR-JAF-2015-05265. The inspectors determined that the failure to inform the NRC of the secondary containment system inoperability within eight hours in accordance with 10 CFR 50.72(b)(3)(v) was a performance deficiency that was reasonably within Entergys ability to foresee and correct. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a safety system functional failure was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that the violation was a SL IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because Entergy personnel failed to make a report required by 10 CFR 50.72 when information that the report was required had been reasonably within their ability to have identified. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects. |
| 05000410/FIN-2016001-04 | Nine Mile Point | 2016Q1 | Licensee-Identified Violation | Eight-hour reports. If not reported under paragraphs (a), (b)(1), or (b)(2) of this section, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the following: (v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material. Contrary to the above, from April 2, 2014, until October 5, 2015, Exelon failed to submit an EN to the NRC within 8 hours upon discovery on a condition which could have prevented the safety function of a SSC needed to control the release of radioactivity on April 2, 2014, at 11:20 a.m. Specifically, secondary containment being declared inoperable due to both airlock doors being open at the same time in Mode 5 with an OPDRV in progress. The inspectors reviewed the violation using IMC 0612 Appendix B, Issue Screening, and the NRC Enforcement Policy. This violation impacted the regulatory process so traditional enforcement applies. Comparing this violation to the examples in the NRC Enforcement Policy Chapter 6, the violation matches Severity Level IV Example 6.9.d.9, a licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. The NRC did not rely upon the information to make any regulatory decisions and the error did not result in increased scope or effort of NRC inspections. Compliance was restored when Exelon submitted LER 05000410/2014-007-01, Secondary Containment Inoperable due to Simultaneous Opening of Airlock Doors, to correct the public record and inform the NRC. Exelon staff entered the issue into its CAP. |
| 05000333/FIN-2016001-03 | FitzPatrick | 2016Q1 | Inadequate Post-Maintenance Testing of the Reactor Building Ventilation System Resulted in Short-Term Inoperability of Secondary Containment | The inspectors identified a self-revealing NCV of TS 5.4, Procedures, for FitzPatrick staffs failure to perform adequate post-maintenance testing (PMT) following maintenance on a limit switch in the reactor building ventilation system in August 2014, that, along with another unrelated component failure in the reactor building ventilation system, resulted in secondary containment pressure, relative to the outside pressure, exceeding the TS limit of 0.25 inches of vacuum water gauge. As immediate corrective action, operators started both trains of the standby gas treatment system (SBGTS), which restored secondary containment pressure to within the TS limit. Operators subsequently secured the A refuel floor exhaust train and placed the B train in service. The issue was entered into the CAP as CR-JAF-2015-04166. The finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, as a result of this event, secondary containment was not preserved, in that secondary containment pressure exceeded the limit of TS surveillance requirement (SR) 3.6.4.1.1. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a pressurized thermal shock issue, did not represent an actual open pathway in the physical integrity of the reactor containment, did not involve an actual reduction in function of hydrogen igniters in the reactor containment, and only represented a degradation of the radiological barrier function provided by the reactor building and SBGTS. The finding had a cross-cutting aspect in the area of Human Performance, Resources, because FitzPatrick staff did not ensure that procedures for PMT of the reactor building refuel floor exhaust damper limit switch following maintenance performed in August 2014, were adequate to support the nuclear safety function of the secondary containment (H.1). |
| 05000461/FIN-2016001-05 | Clinton | 2016Q1 | Failure to Assess and Manage Risk Increase for a Proposed Maintenance Activity | A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65 (a)(4) was identified on January 20, 2016, due to the licensees failure to assess and manage the risk increase from a proposed maintenance activity. Specifically, the licensee failed to manage the risk associated with racking out the continuous containment purge (CCP) A breaker, which resulted in the loss of both CCP trains, and led to an increase in primary to secondary containment differential pressure which exceeded the Technical Specification (TS) value. The licensee entered this issue into their CAP as AR 02614832. The proposed corrective actions to address this issue included creating a checklist to ensure validation of initial conditions is performed and providing training that reinforces the need to properly screen work order tasks with the appropriate risk factors. The inspectors determined that the failure to assess and manage the risk increase of a proposed maintenance activity, as required by 10 CFR 50.65 (a)(4), was more than minor because it was associated with the maintenance procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not properly assessing the risk of racking out the CCP A breaker the licensee did not recognize the CCP B train would be impacted, which resulted in exceeding the TS value for primary to secondary containment differential pressure. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent an actual open pathway in the physical reactor containment, containment isolation system or heat removal components and it did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors identified a cross-cutting aspect in the area of human performance, in the aspect of challenging the unknown, which states, individuals stop when faced with uncertain conditions; risks are evaluated and managed before proceeding. Specifically, when the licensee was preparing the work package for maintenance on the CCP system it was uncertain what activities had already been completed as part of a concurrent evolution. Instead of stopping and validating the configuration of plant equipment, assumptions were made, and the risk of the activity was not properly assessed or managed. (H.11) |
| 05000341/FIN-2016001-07 | Fermi | 2016Q1 | Inadequate Test Criteria in SGTS Flow/Heater Operability Surveillance Test | The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings. Specifically, the licensee failed to include appropriate quantitative or qualitative acceptance criteria in its surveillance test procedures for fulfilling the monthly Technical Specification surveillance requirement to demonstrate operability of the standby gas treatment system (SGTS). The licensee entered this violation into its corrective action program to evaluate the issue and identify appropriate corrective actions. No immediate operability concern was identified. The performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute for the control room and auxiliary building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not providing appropriate acceptance criteria by which the operability of the SGTS trains could be assessed, the ability of the SGTS to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured. The finding was determined to be of very low safety significance because it involved only a degradation of the radiological barrier function provided by the SGTS. The inspectors concluded that because this condition has existed for greater than three years, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified. |
| 05000416/FIN-2015008-03 | Grand Gulf | 2015Q4 | Failure to Declare Secondary Containment Inoperable Based on Failed Surveillance Testing | The team identified a non-cited violation of Technical Specification 3.6.4.1 Condition A, for the failure to declare secondary containment inoperable. Specifically, on August 1, 2015, the licensee failed to declare secondary containment inoperable after it failed to achieve the necessary vacuum to pass Surveillance Requirement 3.6.4.1.4. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2015-05826. The failure to declare secondary containment inoperable due to failed surveillance test and enter the appropriate action statements as required by the licensees technical specifications is a performance deficiency. This deficiency is more than minor, and therefore a finding, because it is associated with the Structures, Systems, Components, and Barrier Performance attribute of the Barrier Integrity cornerstone. Specifically, the failure to declare secondary containment inoperable and take actions as required in Technical Specification Limiting Condition for Operation 3.6.4.1, Condition A, within four hours, adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, dated July 1, 2012, the team determined that the finding is of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the auxiliary building secondary containment. The team determined that this finding has a cross-cutting aspect associated with avoid complacency, in that individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Although the surveillance test was documented as Technical Specification Acceptance Criteria Unacceptable because it did not meet the criteria defined in test procedure 06-OP-1T48-R- 0002, Standby Gas Treatment A Logic and Vacuum Test, Revision 115, the licensee did not identify it as a failed surveillance test that affected secondary containment operability. |
| 05000387/FIN-2015004-02 | Susquehanna | 2015Q4 | Loss of Safety Function of SBGT and CREOASS due to Concurrently Performing Maintenance on Redundant Trains | An NRC-identified finding of very low safety significance (Green) and associated violations of TS 5.4.1, Procedures, TS 5.5.11, Safety Function Determination, and TS 3.7.3, Control Room Emergency Outside Air Supply System was identified when Susquehanna performed maintenance on redundant trains of the standby gas treatment (SBGT) system and control room emergency outside air supply system (CREOASS) concurrently. When performing these actions, operators did not apply NDAP-QA-0312, Control of LCOs, technical requirement for operations (TROs) and Safety Function Determination Program, correctly which resulted in the unrecognized loss of safety function of SBGT and CREOASS. Susquehanna entered the issue into the CAP as CR-2015-26475 and restored one of the subsystems to service, restoring the safety function. 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, allowing work to be performed on redundant trains of SBGT and CREOASS concurrently, while not applying plant TSs correctly, resulted in a loss of safety function of both systems. 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 did not perform a thorough review of the planned activities every time work was performed to ensure compliance with plant TSs, rather than relying on past successes and assumed conditions (H12). |
| 05000352/FIN-2015004-02 | Limerick | 2015Q4 | Licensee-Identified Violation | Technical Specification 3.6.5.3, Standby Gas Treatment System Common System, requires with one SGTS subsystem, restore the inoperable subsystem to operable status within 7 days, or be in at least hot shutdown within the next 12 hours and in cold shutdown within the following 24 hours. Contrary to Technical Specification 3.6.5.3, SGTS subsystem B was inoperable for Unit 1 from August 27, 2015, to September 4, 2015, for a time of 8 days 18 hours, and Exelon did not place Unit 1 in hot shutdown or cold shutdown. Exelon entered this issue into the corrective action program as IR 2517538. 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 finding only represented a degradation of the radiological barrier function of the SGTS. In addition, the inoperable condition would have resulted in a flowrate exceeding the analyzed 2500 cfm with a differential pressure greater than the minimum 0.25 inches of vacuum water gauge. However, the condition did not represent a larger pathway through secondary containment and SGTS retained radiological filtering capability. |
| 05000333/FIN-2015004-02 | FitzPatrick | 2015Q4 | Untimely 10 CFR 50.72 Notification of Inoperable Secondary Containment | The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because inoperability of the secondary containment system was not reported to the NRC within eight hours of when the need to do so should reasonably have been recognized, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. Specifically, positive pressure in the secondary containment due to a previously unidentified equipment malfunction that occurred during transition between the reactor building being isolated and normal reactor building ventilation being in service was not promptly recognized as a condition that caused the single train secondary containment system to be inoperable and therefore to be reportable under 10 CFR 50.72. This issue was entered into the CAP as CR-JAF-2015-05244 and CR-JAF-2015-05265. The inspectors determined that the failure to inform the NRC of the secondary containment system inoperability within eight hours in accordance with 10 CFR 50.72(b)(3)(v) was a performance deficiency that was reasonably within Entergys ability to foresee and correct. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a safety system functional failure was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that the violation was an SL IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because Entergy personnel failed to make a report required by 10 CFR 50.72 when information that the report was required had been reasonably within their ability to have identified. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects. |
| 05000416/FIN-2015008-04 | Grand Gulf | 2015Q4 | Failure to Make Required Event Notification | The team identified two examples of a Severity Level (SL) IV non-cited violation of 10 CFR 50.72(b)(3)(v)(C), for the failure to make an eight-hour report to the NRC for a condition that prevented the fulfillment of the safety function needed to control the release of radioactive material. Specifically, on August 1, 2015, and again on October 1, 2015, after failed secondary containment surveillance tests, the licensee failed to make an eight-hour report to the NRC for the loss of secondary containment barrier safety function needed to control the release of radioactive material. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2015-05826. The failure to report a condition that could have prevented the fulfillment of a systems safety function as required by 10 CFR 50.72(b)(3)(v)(C) is a performance deficiency. This performance deficiency was screened using Inspection Manual Chapter 0612 and was determined to be minor in the Reactor Oversight Process. However, due to the performance deficiency affecting the NRCs ability to perform its regulatory oversight function, this performance deficiency was evaluated for traditional enforcement in accordance with the NRC Enforcement Policy. This performance deficiency was determined to be a Severity Level IV violation in accordance with Section 6.9.d.9 of the NRC Enforcement Policy, dated February 4, 2015. No cross-cutting aspect was assigned to this violation because no Reactor Oversight Process finding exists. |
| 05000341/FIN-2015004-01 | Fermi | 2015Q4 | Failure to Satisfy Technical Specification Requirements During an Unplanned Operation with the Potential to Drain the Reactor Vessel | A finding of very low safety significance with an associated non-cited violation of Technical Specification (TS) 3.0.4 was self-revealed on October 4, 2015, when the licensee inadvertently entered an operation with the potential to drain the reactor vessel (OPDRV) condition while in Mode 5 (refueling) without an operable secondary containment. The licensee failed to provide adequate configuration control of reactor recirculation system boundary isolation valves while establishing conditions to support maintenance during the Cycle 17 refueling outage. As an immediate corrective action, the licensee terminated the OPDRV and restored compliance with the TS by closing recirculation pump seal cavity drain valves to isolate the drain path. In addition, the licensee reviewed all remaining refueling outage system tagouts that interfaced with the reactor vessel to ensure appropriate configuration controls were established to prevent impacting reactor vessel water level, initiated actions to make procedure changes to improve its processes for review of system tagouts for conditions that drain systems that interface with the reactor vessel, and communicated lessons learned from this event with plant operators. The finding was of more than minor safety significance because it was associated with the Configuration Control and Human Performance attributes of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations. Specifically, the system tagout error resulted in an inadvertent and uncontrolled loss of reactor coolant system inventory. The finding was determined to be a licensee performance deficiency of very low safety significance during a detailed Significance Determination Process review since the delta core damage frequency was determined to be less than 1.0E7/year. The inspectors concluded this finding affected the cross-cutting area of human performance and the cross-cutting aspect of avoiding complacency. The cause of the event was primarily attributed to a failure to properly use human error reduction techniques, specifically inadequate self-checking by the operators who prepared and reviewed the system tagout configuration for the maintenance, as well as inadequate identification of OPDRV conditions during refueling outage preparations. |
| 05000237/FIN-2015004-01 | Dresden | 2015Q4 | Failure to Maintain Design Control of Secondary Containment Interlock Doors | A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed on September 4, 2015, when the integrity of the Secondary Containment for Units 2 and 3 was not maintained for 39 minutes when interlock features designed to prevent both doors of a Secondary Containment interlock from being simultaneously open prevented the closure of Reactor Building to Turbine Building doors 47 and 48 following simultaneous operation during routine access of the interlock by plant personnel. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity cornerstone attribute of design control, and adversely affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as very low safety significance (Green) because the inspectors answered yes to the Barrier Integrity Screening Question C.1, Exhibit 3 of IMC 0609, Appendix A. This finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee did not use decision making-practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee failed to implement a modification which addressed a known design deficiency in the 570 foot elevation Secondary Containment interlock in 2013. The licensee reasoned that the interlock was a low traffic area and that it would be unlikely that the doors would be open simultaneously. (H.14) |
| 05000397/FIN-2015003-06 | Columbia | 2015Q3 | Implementation of Enforcement Guidance Memorandum (EGM) 11-003, Revision 2 | During Refueling Outage 22 in May June 2015, Columbia Generating Station implemented the guidance of Enforcement Guidance Memorandum (EGM) 11-003, Revision 2, Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements during Operations with a Potential for Draining the Reactor Vessel, dated December 13, 2013. Consistent with EGM 11-003, Revision 2, secondary containment operability was not maintained during operations with a potential for draining the reactor vessel activities, and required action C.2 of Technical Specification 3.6.4.1 was not completed. The inspectors reviewed this licensee event report for potential performance deficiencies and violations of regulatory requirements. The inspectors reviewed the stations implementation of the EGM 11-003, Revision 2, during operations with a potential for draining the reactor vessel. Specific observations included: 1. The inspectors observed that the operations logged all potential for draining the reactor vessel activities in the control room narrative logs, and that the log entry appropriately recorded the standby source of makeup designated for the evolutions. 2. The inspectors noted that the licensee maintained reactor vessel water level at least greater than 21 feet above the top of the reactor pressure vessel flange as required by Technical Specification 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designed 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. 3. The inspectors verified that the operations with a potential for draining the reactor vessels were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the operations with a potential for draining the reactor vessels. The inspectors verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events. Technical Specification 3.6.4.1, Secondary Containment requires, in part, that secondary containment shall be operable during operations with a potential for draining the reactor vessel. Technical Specification 3.6.4.1, Condition C, requires the licensee to initiate actions to suspend operations with a potential for draining the reactor vessel immediately when secondary containment is inoperable. Contrary to the above, from May 13 - June 13, 2015, Columbia Generating Station performed a total of five operations with a potential for draining the reactor vessel activities while in Mode 5 without an operable secondary containment. These conditions were reported as conditions prohibited by Technical Specifications. The licensee entered this issue into its corrective action program as Action Request 329328. Since this violation occurred during the discretion period described in EGM 11-003, Revision 2, 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 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request (LAR) to resolve the issue for its plant which the NRC staff LAR acceptance review finds acceptable in accordance with LIC-109, Acceptance Review Procedures. The generic solution will be a generic change to the Standard Technical Specifications, and the NRC will publish a notice of availability (NOA) for the TS solution in the Federal Register. Each licensee that receives discretion must submit its amendment request within 12 months of the NRC staffs issuance of the NOA. Licensees may submit LARs to adopt the NRC-approved approach or to propose an alternative approach for their plants. This licensee event report is closed. |
| 05000220/FIN-2015009-01 | Nine Mile Point | 2015Q3 | Failure to Identify and Correct a Condition Adverse to Quality Associated with Secondary Containment Leakage | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion XVI, Corrective Actions, because between 2007 and 2015, Exelon staff did not promptly identify and correct a deficiency associated with Unit 2 reactor building service water pipe penetration W-3177-C. Specifically, on August 20, 2015, during Exelon staffs investigation of an inspector concern associated with the service water pipe penetration into secondary containment, a leakage path into secondary containment was discovered and was not previously identified and evaluated for impact on operability of Unit 2 secondary containment. Exelon generated issue report (IR) 2544831 to document the newly identified condition. The assessment included a review of previously identified leakage paths that were being tracked in accordance with procedure, previously performed secondary containment drawdown leakage testing, and a comparison to the maximum allowable flow rate leakage area. The assessment concluded that based on the gap that was identified at secondary containment penetration W-3177-C, there was a new total of 1.783 square inches of surface area allowing leakage into the Unit 2 secondary containment. Exelon determined this to be acceptable because calculations for secondary containment drawdown testing allows for up to 33.6 square inches of surface area causing in-leakage into secondary containment. Given 1.783 square inches of total identified leakage being less than the allowable 33.6 square inches, there was reasonable assurance that standby gas treatment system will be able to perform its drawdown function and maintain secondary containment vacuum 0.25 inches of vacuum water gauge in accordance with Technical Specification (TS) 3.6.4.1, Secondary Containment. This performance deficiency was more than minor because it impacted the design control attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, Exelons staff failed to identify the degraded penetration seal that impacted the reasonable assurance of Unit 2 secondary containment operability. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined this finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room, or auxiliary, spent fuel pool, or standby gas treatment system (i.e., secondary containment). This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon staff failed to properly evaluate the condition identified in multiple IRs to determine the extent of condition associated with secondary containment water in-leakage. Specifically, between 2007 and 2015, three IRs were generated and a 2012 structural monitoring review documented the service water penetration water in-leakage and the issue was not appropriately evaluated for the potential for a service water pipe through-wall leak or the potential impact on secondary containment. |
| 05000387/FIN-2015003-03 | Susquehanna | 2015Q3 | Secondary Containment Inoperability due to Improperly Controlled Access to the Reactor Building Roof | A self-revealing finding of very low safety significance (Green) and associated NCV of SSES Unit 1 and 2 TS 5.4.1, Procedures, was identified because Susquehanna incorrectly implemented procedures for maintaining secondary containment integrity. Specifically, on July 27, 2015, maintenance technicians rendered secondary containment for both units inoperable for approximately 44 minutes when a secondary containment boundary door was opened to access the reactor building roof. Susquehanna entered the issue into the CAP as CR-2015-20857 and CR-2015-24442, restored the boundary, and verified the integrity of secondary containment. The finding was more than minor because it was 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) protect the public from radionuclide releases caused by accidents or events. Specifically, opening the secondary containment barrier did not maintain reasonable assurance that the secondary containment would be capable of performing its safety function in the event of a reactor accident. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, "The SDP for Findings At-Power," Exhibit 3, for the Barrier Integrity cornerstone, dated June 19, 2012. The inspectors determined the finding was of very low safety significance (Green) because only represented a degradation of the radiological barrier function of secondary containment provided by the standby gas treatment (SBGT) system. This finding had a cross-cutting aspect in the area of Human Performance, Teamwork because Susquehanna did not effectively communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained (H.4). Specifically, when the work plan was changed to accessing the reactor building roof through secondary containment, the change was not effectively communicated to operations department personnel to ensure the secondary containment impairment was appropriately controlled. |