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05000266/FIN-2018002-02Point Beach2018Q2Unanalyzed Condition for Tornado Generated MissilesOn 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 a 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. In the case of Point Beach, 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. Table 1.31 of the Point Beach Final Safety Analysis Report (FSAR) states, in part, that SSCs, which are essential to the prevention and mitigation of nuclear accidents, shall be designed, fabricated, and erected to withstand the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon, such as a tornado. On March 1, 2018, the licensee initiated AR 02252240, identifying a nonconforming condition of Table 1.31. Specifically, on both units 1 and 2, the steam supply lines and exhaust stacks for the turbine-driven auxiliary feedwater pumps, the main steam isolation valves, the atmospheric steam dumps, the main steam safety valves, and the vents for T175B bulk fuel oil storage tank were not adequately protected from tornado-generated missiles. The licensee declared the affected SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice 53239 as an unanalyzed condition and potential loss of safety function. Enforcement discretion was previously authorized and documented in Inspection Report 05000266/2018001 (ADAMS Accession Number ML18128A229). Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of 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. The licensees immediate compensatory measures included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. The licensees longer term compensatory measure was to modify AOP13C, Severe Weather Conditions procedure, to include actions for removing potential airborne hazards and damage assessments for systems with a vulnerability to damage from tornado missiles. Corrective Action Reference: AR 2252240 Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); TS 3.7.1, Main Steam Safety Valves (MSSVs); TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves; TS 3.7.4, Atmospheric Dump Valve (ADV) Flowpaths; TS 3.7.5, Auxiliary Feedwater (AFW); TS 3.8.1; AC Sources Operating; and TS 3.8.3, Diesel Fuel Oil and Starting Air. 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 Number 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 more comprehensive compensatory measures to address the nonconforming conditions within the required 60 days. These comprehensive actions are to remain in place until permanent repairs are completed, which, for Point Beach, 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.On April 26, 2018, the licensee submitted a request to extend the enforcement discretion in letter titled Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15002 for Tornado-Generated Missile Protection Non-conformances Identified in Response to Regulatory Issues Summary 201506, Tornado Missile Protection. On May 21, 2018, the NRC approved this request and extended the enforcement discretion until June 10, 2020. The disposition of this enforcement discretion closes LER 201800100.
05000266/FIN-2018001-04Point Beach2018Q1Enforcement Action: EA18030: Unanalyzed Condition for Tornado Generated MissilesOn 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. In the case of 12 Point Beach, 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. Table 1.31 of the Point Beach Final Safety Analysis Report (FSAR) states in part that SSCs which are essential to the prevention and mitigation of nuclear accidents shall be designed, fabricated, and erected to withstand the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon such as a tornado. On March 1, 2018, the licensee initiated AR 02252240, identifying a nonconforming condition of Table 1.31. Specifically, on both units 1 and 2, the steam supply lines and exhaust stacks for the turbine-driven auxiliary feedwater pumps, the main steam isolation valves, the atmospheric steam dumps, the main steam safety valves, and the vents for T175B bulk fuel oil storage tank were not adequately protected from tornado-generated missiles. The licensee declared the affected SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice (EN) 53239 as an unanalyzed condition and potential loss of safety function. Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of 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. The licensees immediate compensatory measures included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. Corrective Action Reference: AR 2252240 Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); TS 3.7.1, Main Steam Safety Valves (MSSVs); TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves; TS 3.7.4, Atmospheric Dump Valve (ADV) Flowpaths; TS 3.7.5, Auxiliary Feedwater (AFW); TS 3.8.1; AC Sources - Operating; and TS 3.8.3, Diesel Fuel Oil and Starting Air. 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 15-002 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 are to remain in place until permanent repairs are completed, which for Point Beach 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
05000364/FIN-2018001-02Farley2018Q1Enforcement Action (EA)-18-025:Unit 2 Main Steam Safety Valve (MSSV) Lift Pressure Outside of Technical Specification LimitsOn October 26, 2017, MSSV Q2N11V0012E was removed from service at Farley Nuclear Plant Unit 2 during a refueling outage, and on November 1, 2017 the valve was tested with steam at an offsite facility. As-found lift testing determined that the valve opened at 1171 psig steam pressure, which was 9 psig high outside the plant technical specification (TS) allowable lift setting range of 1096 psig to 1162 psig. The valve had been in service prior to the plant beginning commercial operation on July 30, 1981, until it was removed from the main steam system on October 26, 2017. The licensee last tested the valve, while installed on the main steam system, on April 5, 2016. The test results indicated the lift pressure was within +/- 1% of the TS 3.7.1 required set pressure of 1129 psig, and no set pressure adjustment was necessary for the valve. The licensee determined that the MSSV high as-found lift set-point did not have an adverse impact on the main steam system over-pressurization protection, since the valve as-found lift setpoint was lower than 110% of steam generator design pressure (1194 psig), and this condition would not have resulted in a loss of safety function. Therefore, the plant remained bounded by the accident analysis in the Final Safety Analysis Report (FSAR), based on the as-found condition. Corrective Action(s): The valve was replaced with an operable MSSV during the refueling outage prior to plant startup.Corrective Action Reference(s): The licensee entered this issue into their Corrective Action Program (CAP) as condition report (CR) 10426186 as found test results for MSSV Q2N11V0012E. Violation: Farley Nuclear Plant, Unit 2 Technical Specifications (TS) limiting condition for operation (LCO) 3.7.1, Main Steam Safety Valves (MSSVs), required five MSSVs per steam generator to be operable. Per TS Table 3.7.1-2, MSSV Q2N11V0012E must have a lift setting within the range of 1096 psig to 1162 psig, while the Unit was in modes 1, 2, and 3. With one MSSV inoperable and the Moderator Temperature Coefficient (MTC) zero or negative at all power levels, Action Statement, Condition A, Required Action A.1, required reducing thermal power to 87% RTP within 4 hours. If the required action and associated completion time is not met, Action Statement, Condition C, required that the unit be in mode 3 within 6 hours.Contrary to the above, the licensee determined the MSSV setting was outside the TS limits longer than 10 hours during the operating cycle between May 11, 2016 and October 15, 2017, while the Unit was in modes 1, 2, and 3. Severity/Significance: The inspection assessed the severity of the violation using Section 6.1 of the Enforcement Policy and determined the significance is appropriately characterized at Severity Level IV, due to the inappreciable potential safety consequences. The significance of this violation was informed, in part, using IMC 0609, Appendix A, The Significance Determination Process (SDP) for findings at Power, dated June 19, 2012. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section3.10 of the Enforcement Policy because the MSSV as-found lift pressure issue was not reasonably foreseeable and preventable. The inspectors reached this conclusion due to the fact that the licensee last tested the valve satisfactorily, while installed on the main steam system, on April 5, 2016, and during the period of time that the valve was in service, following May 11, 2016, there was no indication of valve degradation (e.g. seat leakage)
05000454/FIN-2017010-01Byron2017Q4Failure to Prevent Secondary Missiles Following a Postulated HELBThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regualtions (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design basis for the main steam safety valve (MSSV) room maintenance hatches was maintained. Specifically, the high energy line break ( HELB) analysis performed for the MSSV rooms and steam tunnels prior to initial construction concluded that no secondary missiles were generated as a result of a HELB although maintenance hatches in the ceiling of the MSSV rooms were identified to become secondary missiles following a HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their corrective action program (CAP) as AR 4075608 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of system s that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC (Structure, System, and Component) , does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross- cutting aspect was assigned to this finding as it was not reflective of current performance.
05000424/FIN-2016007-08Vogtle2016Q4Failure to Promptly Identify Nonconformances with Tornado Missile ProtectionEnforcement Guidance Memorandum (EGM) 15-002 dated 6/10/2015, (ADAMS Accession No. ML15111A269) provided guidance to exercise enforcement discretion when an operating power reactor licensee does not comply with the plants current site-specific licensing basis for tornado-generated missile protection. Specifically, discretion would apply to the TS limiting conditions for operation (LCO) which would require a reactor shutdown or mode change, if a licensee could not meet TS LCO required action(s) within the TS completion time. The EGM background discussed Regulatory Issue Summary (RIS) 2015-06, Tornado Missile Protection, dated 6/10/2015, (ADAMS Accession No. ML15020A419) to remind licensees of the need to conform their facility to the current, site-specific licensing basis for tornado-generated missile protection. In addition the EGM stated, that upon reviewing the above-noted RIS, some licensees may discover that a TS-controlled SSC at their facility does not comply with the plants current licensing basis (CLB) and that an operability determination (or functional assessment) will be necessary. The EGM actions section specified that the NRC would exercise this enforcement discretion only when a licensee implements initial compensatory measures prior to the expiration of the time allowed by the LCO that provide additional protection such that the likelihood of tornado missile effects are lessened. The licensee initiated CR10087558 on 06/23/2015, to evaluate the RIS and conducted at least two walk-downs to identify tornado missile nonconformances. The licensee discovered potential nonconformances during these walk-downs and itemized them in a list. However, the licensee failed to identify all of these items as conditions adverse to quality (CAQs), in accordance with Appendix B, Criterion XVI. The team determined that the CAP required the evaluation of these items, CRs to document the nonconformances, and operability determinations for items affecting TS. Procedure NMP-GM-002, Corrective Action Program, Section 2, defined a condition adverse to quality in part, as an all-inclusive term used in reference to any of the following: ..., deficiencies, ..., and nonconformances potentially impacting Nuclear Safety. Nonconformances are deficiencies in characteristic, documentation, or procedure that renders the quality of an item or activity unacceptable or indeterminate. The team determined that, at the time of discovery, the itemized tornado missile vulnerabilities rendered the quality of SSCs indeterminate and thus a nonconformance in accordance with the definition in the procedure. Procedure NMP-GM-002-001, Corrective Action Program Instructions Section 4 specified that personnel should initiate a CR to identify an event, condition, problem, or process that needs correcting. (This included) nonconforming items. In addition, Section 4 specified to immediately contact the Shift Support Supervisor or Work Week Coordinator (Dispatcher) when a condition is discovered that has the potential to impact plant operation or reportability. (This included) equipment or process issues related to Technical Specifications (tech specs). The team noted that the licensee did not create any additional CRs for the itemized potential vulnerabilities as required by their corrective action instructions procedure. On October 4, 2016, the inspectors conducted plant walk downs of the SSCs selected in the CDBI inspection plan and identified potential tornado missile issues. These issues were previously highlighted as potential nonconformances by the licensee, but not identified as CAQs. As a result of these observations, the licensee initiated CRs: CR10291142, Unit 1 TDAFW Exhaust nonconformance CR10291143, Unit 2 TDAFW Exhaust nonconformance CR10291144, Unit 1 Condensate Storage Tanks nonconformance CR10291145, Unit 2 Condensate Storage Tanks nonconformance CR10291146, Unit 1 Main Steam Safety Valve Exhaust nonconformance CR10291148, Unit 2 Main Steam Safety Valves Exhaust nonconformance The licensee determined that the TDAFW Exhaust and Condensate Storage Tanks were not operable because of nonconformances with these components tornado missile protection design bases. Additionally, the licensee submitted a 10 CFR 50.72 notification report (52319) to the NRC in accordance with plant procedures and NRC requirements.
05000286/FIN-2015002-01Indian Point2015Q2Inadequate Corrective Action for Main Steam Safety Valve 46-3 Failure to Lift at Required SetpointThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergys failure to take corrective actions for a condition adverse to quality involving Unit 3 Main Steam Safety Valve (MSSV) 46-3. Specifically, MSSV 46-3 failed to meet its Technical Specification (TS) required lift setting during a surveillance test on March 22, 2015. This failure was documented in a condition report (CR) but closed for trending purposes. Additionally, Entergy personnel did not correct the failure of MSSV 46-3 to meet its TS required lift setting after it failed its as-found lift setting test on March 1, 2013. The inspectors determined the performance deficiency was more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, Entergy did not take corrective actions following the March 22, 2015, failure of MSSV 46-3, and previous corrective actions in 2013 were not effective in ensuring it would remain capable of lifting at its TS required setpoint. The inspectors determined that this finding is of very low safety significance (Green) because the finding does not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with Entergys maintenance rule program for greater than 24 hours. Specifically, of the 20 valves tested in 2015, 16 passed the as-found lift test and there was no loss of safety function. The inspectors determined that this finding had a Problem Identification and Resolution cross-cutting aspect related to Evaluation, because Entergy did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the CR documenting the MSSV 46-3 failure was closed for trending purposes and as a result, a thorough evaluation of the cause was not completed.
05000286/FIN-2015002-04Indian Point2015Q2Licensee-Identified ViolationOn February 27, 2015, MSSV 46-2, 45-4, and 47-4 failed to meet the TS as-found lift setpoint test. Additionally, on March 22, 2015, MSSV 46-3 failed to meet the TS asfound lift setpoint test. TS Limiting Condition for Operation 3.7.1.A, Main Steam Safety Valves, requires if one or more required MSSVs are inoperable, reduce neutron flux trip setpoint to less than or equal to the applicable percent Reactor Thermal Power listed in Table 3.7.1- 1 within 4 hours. If the required action and associated completion time is not met, the reactor shall be placed in Mode 3 in 6-hours and Mode 4 in 12 hours. Contrary to this requirement, MSSV 46-2, 45-4, 47-4, and 46-3 were inoperable for a time period that exceeded the TS allowed outage. This finding was determined to be of very low safety significance (Green) because the finding did not represent a loss of safety function for the MSSV system. This issue was documented in Entergys CAP (CR-IP3-2015-0898 and CR-IP3-2015-2128) and reports were made to the NRC in LER 05000286/2015-002-00 and 2015-002-01.
05000286/FIN-2015002-03Indian Point2015Q2Incomplete 50.73 Report Associated with Failures of Main Steam Safety ValvesThe inspectors identified a Severity Level IV NCV of 10 CFR 50.9(a); in that, Entergy did not provide complete information in a report submitted per 10 CFR 50.73(a)(2)(i)(B). Specifically, a Licensee Event Report (LER) submitted on April 27, 2015, which reported three MSSV test failures (MS-46-2, MS-45-4, MS-47-4) that occurred on February 27, 2015, did not discuss the failure of MSSV 46-3, which also failed its TS as-found lift setting test and was declared inoperable on March 22, 2015. MSSV 46-3 was inoperable for greater than its TS allowed outage time, which is a condition prohibited by TSs, and therefore is required to be reported to the NRC. The inspectors evaluated this performance deficiency in accordance with the Traditional Enforcement process. In accordance with Section 2.2.2.d of the NRC Enforcement Policy, the inspectors determined that the performance deficiency identified with the reporting aspect of the event is a Severity Level IV violation, because it is of more than minor concern, with relatively inappreciable potential safety significance and is related to findings that were determined to be more than minor issues. Specifically, this issue is related to a more than minor corrective action finding, which is documented in Section 1R22 of this report. In accordance with IMC 0612, Appendix B, this traditional enforcement issue is not assigned a cross-cutting aspect.
05000400/FIN-2015001-01Harris2015Q1Licensee-Identified ViolationTS 6.8, Procedures and Programs, Section 6.8.1.a requires, in part, that written procedures be established, implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide (RG) 1.33, Revision 2, February 1978. RG 1.33, Appendix A, Section 8.b.(1).(dd) requires, in part, that procedures be established for safety valve surveillance tests. Contrary to the above, on March 12, 2015, engineers used an inadequate procedure to test main steam safety valves. Specifically, engineering procedure EST-224, Insitu Main Steam Safety Valve Test using Assist Device, did not adequately direct personnel to compensate for the head differential between the pressure gauge and the seat of the main steam safety valves. This resulted in the licensee incorrectly adjusting the setpoint of MSSV MS-46 below TS 3.7.1 limits while operating in Mode 1. The licensee identified this issue after incorrectly declaring the valve operable. However, the licensee was able to restore the setpoint and operability of MS-46 within the TS 3.7.1 Limiting Condition for Operation action time. This violation was determined to be of very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The licensee entered this issue into their CAP as AR #737961. As corrective actions, the licensee revised the procedure and restored the setpoint to within TS 3.7.1 limits and retested MS-46.
05000286/FIN-2014005-07Indian Point2014Q4Licensee-Identified ViolationOn March 1, 2013, Entergy personnel tested Unit 3 main steam safety valves and determined main steam safety valve MS-46-3 had a lift setpoint outside of the +/-3 percent lift setting required by TS 3.7.1. Subsequently, MS-46-3 was declared inoperable and further testing found valve MS-48-3 also lifted out of the TS band. TS 3.7.1 requires the main steam safety valves be operable or reduce neutron flux trip setpoint to less than that listed in TS Table 3.7.1-1. Contrary to the above, as of March 1, 2013, main steam safety valves MS-46-3 and MS-48-3 had lift setpoints outside of the TS required band and flux trip setpoints were not reduced to those listed in TS Table 3.7.1-1. The affected valves were adjusted at the time of testing to within the required band, the condition was documented in the CAP as CR-IP3-2013- 0869 and CR-IP3-2013-0892, and an evaluation was initiated. Other valves similarly tested were satisfactory. No performance deficiency was identified because it was not reasonable for Entergy to foresee and prevent the change in main steam safety valve setpoint during plant operation. Corrective actions to prevent recurrence were documented in LER 05000286/2013-001-00. The violation was more than minor because it impacted the Equipment Performance attribute of the Mitigating Systems cornerstone. The issue screened to be of very low safety significance (Green) using IMC 0609, Appendix A because the overall pressure mitigating function was not affected by the degradation of the two valves of the twenty total.
05000275/FIN-2014004-04Diablo Canyon2014Q3Inadequate Procedure Results in Unnecessary Main Steam Safety Valve LiftThe inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensee failure to prescribe a procedure appropriate to the circumstances with respect to safetyrelated atmospheric dump valves and main steam safety valves. Specifically, control of atmospheric steam dump valves was not appropriate for a rapid plant shutdown resulting in unnecessary lifting of a spring-loaded main steam safety valve. The inspectors determined that the licensees failure to ensure appropriate procedures to properly control steam generator pressure and prevent unnecessary lifting of main steam safety valves was a performance deficiency. This performance deficiency was determined to be more than minor because it affected the Mitigating Systems cornerstone attribute of procedural quality and the objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component that did not affect operability or functionality. The inspectors concluded that this finding affected the cross-cutting aspect of human performance associated with avoiding complacency, because the licensee failed to recognize during rapid load reductions the inherent risk of lifting a main steam safety valve and did not recognize or plan with adequate procedures, for a condition with a potential latent problem.
05000400/FIN-2014007-03Harris2014Q3Failure to Establish Appropriate Procedural Limitations to Prevent Exceeding TS Limits and Safety Analysis AssumptionsThe team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that applicable regulatory requirements in technical specification (TS) 3.7.1.1 and design basis inputs in accident analyses were translated into procedural guidance. Specifically, the licensee did not follow their inservice test program guidance to account for surveillance test equipment instrument uncertainty when establishing the acceptability of Main Steam Safety Valve lift setpoints required by TS 3.7.1.1. Following identification by the team, the licensee generated nuclear condition report 697100 and performed an evaluation of the remaining available margin to the overpressure limit in the safety analysis, and discovered that, after potential instrument uncertainty was taken into account, the margin remained positive, but was reduced from approximately 19 psig to approximately 6 psig. The licensees failure to assure that applicable regulatory requirements in TS 3.7.1.1 and design basis assumptions in accident analyses were correctly translated into procedural guidance, as required by 10 CFR Part 50, Appendix B, Criterion III, was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, by not accounting for the measurement and test equipment uncertainties as required by the inservice test program, it could have led to the actual lift setpoints exceeding the inputs used in the design basis safety analyses. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance.
05000445/FIN-2013003-01Comanche Peak2013Q2Inadequate Procedure for Testing the Main Steam Safety ValvesThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of the licensee to have documented instructions of a type appropriate to the circumstances for testing the main steam safety valves. Specifically, the procedure for testing the main steam safety valves did not provide direction to declare the valves inoperable when applying pressure to the lifting device. As a result, the licensee failed to declare the main steam safety valves inoperable during testing. The licensee entered the finding in the corrective action program as Condition Report CR-2013-002947. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedure did not provide guidance to declare a main steam safety valve inoperable with the test rig installed. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance because the finding was not a design or qualification deficiency; did not represent an actual loss of safety function of a system or train; and did not result in the loss of one or more trains of non-technical specification trains of equipment. The inspectors determined that the finding was not representative of current licensee performance and no cross-cutting aspect was assigned.
05000285/FIN-2013008-27Fort Calhoun2013Q2Continuous Monitoring Capability of Post Accident Main Steam Radiation Monitor RM-064The team identified an unresolved item associated with post accident radiation monitor RM-064. Specifically, the team is concerned about the capability of the monitor to provide representative measurements due to the system configuration, and this could represent a failure to ensure continuous effluent monitoring of the main steam lines following a steam generator tube rupture accident. Following the Three Mile Island Accident in March 25, 1979, licensees were required to ensure all potential effluent release points from nuclear power plants were equipped with high range radiation monitors. In particular, NUREG-0737, Section II.F.1.1 requires in part; that for pressurized water reactors such as FCS, Unit 1, steam release points be monitored for noble gases, and that indication of the activity must be monitored and recorded continuously. In addition Section II.F.1.1 requires the monitors shall be capable of functioning both during and following an accident. System designs shall accommodate a design-basis release and then be capable of following decreasing concentrations of noble gasses. In addition, the monitoring system shall be capable of obtaining readings at least every 15 minutes during and following an accident. . The team identified an unresolved item associated with post accident radiation monitor RM-064. Specifically, the team is concerned about the capability of the monitor to provide representative measurements due to the system configuration, and this could represent a failure to ensure continuous effluent monitoring of the main steam lines following a steam generator tube rupture accident. By application dated March 9, 1984, the licensee requested an amendment to the stations technical specifications in response to the Commissions Generic Letter 83-37, NUREG-0737 Technical Specifications. The generic letter, which was issued in November 1, 1983, advised licensees to submit new technical specifications for NUREG-0737 items, including Section II.F.1.1, Noble Gas Effluent Monitors (II.F.1.1). The stations potential post-accident steam release points include the main steam relief valves, the atmospheric dump valve, and the steam driven AFW pumps steam turbine. To comply with the high range radiation monitoring requirements, the licensee installed noble gas effluent monitors including, radiation monitor RM-064. Per USAR Section 11.2.3.11, RM-064, the post-accident main steam line monitor, is an off-line monitor designed to measure the steam activity by sampling steam from the two steam headers via two isolation valves HCV-921 and HVC-922. The monitor is placed in service in the event of a steam generator tube rupture. The monitor is capable of sampling steam from both steam headers and the recorded data from this monitor can then be utilized to quantify effluents released through the main atmospheric dump valve, the main steam safety valves, and the AFW pump turbine. Radiation monitor RM-064 is located in the turbine building next to Room 81. - 223 - The team noted that the design basis accident analysis contained in USAR, Section 14.14, Steam Generator Tube Rupture Accident, required the licensee to assume a coincident reactor trip and a loss of off-site power. Due to the assumed simultaneous loss of off-site power with the reactor trip, the reactor is cooled down by releasing steam via the main steam safety valves and atmospheric dump valve, creating a direct release path to the environment. In addition, due to the loss of off-site power, the normal condenser off-gas radiation monitor becomes un-available due to the loss of condenser vacuum. This leaves radiation monitor RM-064 as the only monitor available to measure radioactivity in the main steam lines. The analysis assumes all activity released from the faulted steam generator ceases when it is isolated by plant operators 2 hours after the event. The design of the FCS main steam line monitor is provided in MR-FC-79-190C, Post Accident Main Steam Line High Range Radiation Monitor RM-064, Revision 0, dated June 4, 1982. The station has two 28 inch diameter headers leading to the main turbine. Each main steam line is provided with six main steam safety valves each having different lift set-points. The pipe connecting these valves is 2.5 inches in diameter. The pipe connecting to the atmospheric dump valve is 3 inches in diameter. The sample line to radiation monitor RM-064 is 3/8 inch in diameter. This line is located upstream of the main steam isolation valves, in Room 81 of the auxiliary building. The distance from the main steam header to the actual location of radiation monitor RM-064 (outside Room 81) is over sixty feet long, while the main steam safeties and steam dump valve, are within 12 feet away from the main steam headers. The team reviewed the USAR, main steam drawings, applicable calculations, and interviewed engineers and operators to identify the design basis requirements for radiation monitor RM-064 and to verify it was capable of performing its intended functions. On The team also determined that for the B steam generator header the location of the 3/8 inch sample line leading to radiation monitor RM-064 was installed downstream of three of the main steam safety valves, including the lowest lift set-point valve. For the A steam generator header, the 3/8 inch sample line was located downstream of two of the safety valves but upstream of the lowest lift set point relief. Due to the location of the sample lines being downstream of the safety valves, the difference in pipe sizing between the lines to the monitor (3/8 inch), the main steam safety valves (2.5 inch), and the atmospheric dump valve (3 inch) and the distance from the main steam header to the monitor, the team questioned how the licensee assured a representative measurement would be obtained during and after a steam generator tube rupture accident. The team informed the licensee of their concerns and the licensee initiated Condition Reports CR 2013-04442, 2013-05515, and 2013-06267, to capture these concerns in the CAP. February 27, 2013, the team performed a walkdown of radiation monitor RM-064 and the steam lines. Because radiation monitor RM-064 is normally isolated, the team questioned how long it would take operators to put the monitor in service, and how the licensee met the requirement of continuous monitoring. During subsequent evaluations the licensee determined that there was not an established time requirement for operators to put radiation monitor RM-064 in service. - 224 - The licensee performed a simulator dry run with licensed operators to estimate the time required to place the monitor in service. During this simulated event, it took operators approximately 23 minutes to put the monitor in service, thus indicating that there could be an unmonitored release to the environment for at least 23 minutes following a steam generator tube rupture accident. Regarding the representative sample concern, engineers determined that without a sophisticated computer model it could not be definitely shown that the degree of turbulent mixing in the steam lines is sufficient to equalize the concentrations of radioactive gasses and entrained particulates downstream of the main steam safety valves where the lines connecting to radiation monitor RM-064 were located. The licensee issued Condition Report CR 2013-10507 requesting a detailed calculation to address this concern. The team determined this condition has existed since the time radiation monitor RM-064 was installed in February 1983, until February 27, 2013, when the issue was identified by the team. An engineering technical evaluation was then performed under Condition Report CR 2013-04442, based on existing radiological analysis Calculation FC06820 used for the steam generator accident analysis (USAR 14.14). This technical evaluation removed many of the conservative assumptions included in Calculation FC06820. Based on this basic evaluation and using engineering judgment, the licensee determined that there would be sufficient mixing and adequate concentration to provide a representative radiation measurement. The team concluded that further review is necessary in order to properly evaluate and disposition this issue. This issue is identified as URI 05000285/2013008-27, Continuous Monitoring Capability of Post Accident Main Steam Radiation Monitor RM- 064.
05000247/FIN-2012003-02Indian Point2012Q2An LER for an Inoperable Main Steam Safety Valve Was Not Submitted When RequiredThe inspectors identified a Severity Level lV, NCV of 10 CFR 50.73(a)(2)(i)(B), because Entergy personnel did not provide a written licensee event report (LER) to the NRC within 60 days of identifying during testing that MS-46D, main steam line safety valve, was inoperable and in a condition prohibited by the plants Technical Specification (TS). Entergy personnel adjusted the valves lift setpoint to within the TS operability limit, repaired and tested the valve before plant startup. Entergy staff entered this issue into the CAP as CR-IP2-2012-3320 and CR-IP2-2012-4153. The inspectors determined that the failure to provide a written LER within 60 days was a performance deficiency that was reasonably within Entergys ability to foresee and correct, and should have been prevented. This violation involved not making a required report to the NRC and is considered to impact the regulatory process. Such violations are dispositioned using the traditional enforcement process instead of the Significance Determination Process. Using the NRC Enforcement Policy Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, example (d)(9), the NRC determined this violation is more than minor and is categorized as a Severity Level IV violation. Because this violation involves the traditional enforcement process with no underlying technical violation that would be considered more than minor in accordance with IMC 0612, a cross-cutting aspect is not assigned to this violation.
05000336/FIN-2011003-03Millstone2011Q2Inadequate Corrective Action Results in Loss of Enclosure Building\'s Safety FunctionA self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XVl, Corrective Action, was identified for Dominion\'s failure to take prompt corrective action to address the cause of main steam safety valve (MSSV) exhaust pipe bushings not seating, which resulted in a loss of the Enclosure Building\'s safety function to control the release of radioactive material. Dominion took corrective action to clean and lubricate the MSSV exhaust pipe and also implemented a modification to upgrade the MSSV outlet boot and qualify it as part of the Enclosure Building filtration boundary (cR420485). The finding was more than minor because it was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure of the MSSV sliding bushings to seat properly caused the Enclosure Building Filtration System (EBFS) to fail its surveillance test, and its safety function to control the release of radioactive material could not be assured. The inspectors conducted a Phase 1 screening in accordance with NRC Inspection Manual Chapter (lMC) Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it only represents a degradation of the radiological barrier function provided for the auxiliary building. The inspectors determined that this finding had a cross-cutting aspect in the Problem ldentification and Resolution cross-cutting area, Corrective Action Program component, because Dominion did not take appropriate corrective action to address the Enclosure Building surveillance test failure in 2009.
05000289/FIN-2010009-03Three Mile Island2010Q3Multiple MSSVs test failures due to improper evaluation of Operating ExperienceA self-revealing Green NCV of TMI Technical Specification (TS) 3.4.1.2.3 was identified for having greater than three main steam safety valves (MSSVs) inoperable for greater than the allowed outage time with reactor power greater than 5%. MSSV testing prior to the 2009 refueling outage identified that six MSSVs failed the lift point test and were subsequently declared inoperable. All six valves failed by lifting above the ASME limit of +/- 3% of designed setpoint. Five of these six valves exhibited signs of oxide binding, a known failure mechanism for MSSVs and each of the valves had been refurbished during the 2007 refueling outage. Exelon had fleet and industry information about the oxide binding failure mechanism available in 2006 at the time the refurbishment method was selected for the 2007 TMI outage. This refurbishment method included a decision to machine hone the MSSV seat to a mirror finish. This decision created the conditions for oxide binding and resulted in each of the valves failing their lift tests and being declared inoperable when tested in 2009. Exelon has changed its refurbishment process to preclude this error in the future, refurbished all of the affected valves, submitted a required licensee event report (LER), and entered the issue into the CAP. The decision in 2006 to machine hone the MSSV seat to a mirror finish, which established the conditions for oxide binding, was a performance deficiency that was within Exelon\'s ability to foresee and prevent due to available operational experience. This performance deficiency is more than minor because it affected the Equipment Performance Aspect of the Mitigating Systems Cornerstone Objective of ensuring the operability, availability, and reliability of systems designed to mitigate transients and prevent core damage. The team assessed this finding in accordance with IMC 0609, Attachment 4, Phase 1 - \"Initial Screening and Characterization of Findings,\" and determined that it was of very low safety significance (Green) since it did not result in a loss of any system safety function. This finding was determined to not have a cross-cutting aspect because the performance deficiency occurred in 2006 and was no longer indicative of current licensee performance. Specifically, Exelon made changes to their MSSV refurbishment program in 2008 which implemented the available OE, prior to discovery of the 2009 failures.
05000247/FIN-2010003-02Indian Point2010Q2Licensee-Identified ViolationTS 3.7.1, main steam safety valves (MSSVs), requires that MSSVs shall be operable, which, in part, is specifically met if as-found lift setpoints are within applicable acceptance criteria during in-service testing. Contrary to this requirement, on March 9, 2010, during performance of MSSV testing, Entergy personnel identified that MS- 45C and MS-48C exceeded as-found lift set pOints. Entergy technicians subsequently performed satisfactory adjustments and as-left testing to ensure operability was restored. Entergy documented this issue in the corrective action program for resolution under condition report CR-IP2-2009-01181. In addition, Entergy personnel analyzed the past operability and associated impact on the safety analysis with two MSSVs potentially lifting at greater than allowable setpoints and concluded that the condition would not have prevented the accident mitigation capability of the MSSVs overpressure function. Although two MSSVs were determined to be inoperable for an unknown duration, and potentially longer than the allowed outage time listed in the Unit 2 technical specifications, the inspectors determined that this finding is of very low safety significance because it did not increase the probability or consequences of any anticipated operational occurrence or accidents covered by the safety analysis.
05000317/FIN-2010003-02Calvert Cliffs2010Q2Inadequate Design Control Reviews of the Turbine Control System and the Nuclear Steam Supply SystemThe inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, \\\"Design Control,\\\" because Constellation did not perform adequate design reviews associated with modifications to the turbine control system and the nuclear steam supply system (NSSS). Specifically, Constellation did not adequately evaluate the potential adverse impacts of removal of the power load unbalance (PLU) turbine trip on safety related systems, structures, and components (SSCs) such as the main steam safety valves (MSSVs) and pressurizer power operated relief valves (PORVs). In addition, during significant changes to plant design such as steam generator replacements and power uprates, Constellation did not conduct an adequate evaluation to determine if the turbine bypass valve (TSV) and the atmospheric dump valve (ADV) design specification of opening within 3 seconds after receiving the quick open signal would still be sufficient to prevent lifting MSSVs. Immediate corrective actions included entering these issues into their corrective action program (CAP) and performing an immediate operability determination and a probabilistic risk analysis. This finding is more than minor because it affected the Initiating Event cornerstone attribute of design control and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the removal of the PLU turbine trip and the modifications to the NSSS could challenge prirnary and secondary overpressure protection devices and result in a stuck open MSSV or PORV. The inspectors evaluated this finding using an SDP phase 2 analysis and deterrnined that the issue is of very low safety significance. This finding has a cross-cutting aspect in the area of hurnan performance, decision rnaking, because Constellation did not adequately make safetysignificant decisions using a systematic process when faced with uncertain or unexpected plant conditions, to ensure safety is maintained. (H.1.a of IMC 0310).
05000286/FIN-2010002-03Indian Point2010Q1Licensee-Identified ViolationTS 3.7.1 requires that all main steam safety valves (MSSVs) shall be operable, which, in part, is specifically met if as-found lift setpoints are within applicable acceptance criteria during in-service testing. Contrary to this requirement, on March 10, 2009, during performance of MSSV testing, Entergy personnel identified that MS-45-1 and 48-3 exceeded as-found lift setpoints. Entergy subsequently performed satisfactory adjustments and as-left testing to ensure operability was restored. Entergy documented this issue in the corrective action program for resolution under condition report CR-IP3-2009-00716. In addition, Entergy personnel analyzed the past operability and associated impact on the safety analysis with two MSSVs potentially lifting at greater than allowable setpoints. Although two MSSVs were determined to be inoperable for an unknown duration, and potentially longer than the allowed outage time listed in Unit 3 technical specifications, the inspectors determined that this finding is of very low safety significance because it did not increase the probability or consequences of any anticipated operational occurrence or accidents covered by the safety analysis.
05000445/FIN-2008005-06Comanche Peak2008Q4Failure to Have an Adequate Procedure to Test Main Stream Safety ValvesThe inspectors documented a self-revealing noncited violation of Technical Specification 5.4.1a (Procedures) for an inadequate test procedure that resulted in inadvertently holding open a main steam safety valve at power. During testing, a test engineer separated a quick disconnect fitting in accordance with the procedural instructions. The action sealed in nitrogen pressure in the test rig and caused the valve to remain held open. In response to the event, operators reduced reactor power to compensate for the partially open safety valve until maintenance personnel closed the valve. The licensee entered the finding into their corrective action program as Smart Form SMF-2008-002946. The finding was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone, and directly affected the cornerstone objective to limit the likelihood of those events that upset plant stability during power operations. Using Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding had very low safety significance because it did not contribute to the likelihood of mitigating equipment being unavailable. This finding did not have a crosscutting aspect because the procedure section was last revised several years earlie
05000456/FIN-2008002-02Braidwood2008Q1Licensee-Identified ViolationTechnical Specification 3.7.1 required that the plant be in Mode 4 within 12 hours when any one steam generator had four or more main steam safety inoperable. Contrary to this, between October 24, 2007 and October 26, 2007, both the A and D steam generators operated in Modes 1, 2, and 3 with 5 inoperable main steam safety valves each. This was identified in the licensees corrective action program as Issue Report 689141. This finding was of very low safety significance because it did not represent a loss of secondary plant heat removal capability or secondary plant depressurization capability
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