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| Title | Quarter | Description | |
|---|---|---|---|
| 05000334/FIN-2000005-02 | Problem Identification and Resolution | 2000Q3 | No Color On two occasions, problem assessments did not properly evaluat potential risk significance and implement timely effective corrective actions Although these deficiencies were not the root or contributing causes to the actua events, they represent adverse performance which limited the licensees ability t identify and correct adverse safety conditions. Specifically, 1) station personne did not recognize the potential risk significance of the degraded A auxiliar river water pump seal and did not correct the condition in a timely manner; an 2) the safety significance assessment for a reactor protection system (RPS miscalibration event was also deficient, in that engineers incorrectly conclude that protective functions of the instrument channel were not affected Additionally, corrective actions for the RPS miscalibration event did not preclud two repeat miscalibration occurrences. |
| 05000334/FIN-2002003-01 | Failure to Adequately Test or Maintain the Sirens (Phads) to Meet the Original Design Basis of the ANS | 2002Q1 | The NRC performed the supplemental inspection to assess the licensees evaluation and corrective actions regarding the inadequate maintenance and testing of the personal home alerting devices (PHADs), an integral part of the alert and notification system, to ensure that the design function of alerting essentially 100% of the public in the emergency planning zone could be met. This performance issue was previously characterized as having low to moderate safety significance (i.e. White) in NRC Inspection Report 50-334/02-003 and 50-412/02-003. During this supplemental inspection, performed in accordance with Inspection Procedure 95001, the inspector determined that the licensee performed a comprehensive evaluation of the inadequate PHAD maintenance and testing. First Energys evaluation identified the primary root cause of the performance issue to be the failure to administratively ensure that PHAD test procedures, acceptance criteria, maintenance programs and installation data were properly established and maintained. However, First Energy performed an extent of condition review which was not o sufficient scope, in that it focused on EP issues alone. Additionally, the effectiveness review, completed to assess the effectiveness of corrective actions, focused solely on the siren hardware changes and did not cover the human performance issues which contributed to inadequate PHAD testing and maintenance. Subsequent NRC review of a new extent of condition investigation documented in CA #24 to CR 02-02202 and effectiveness review documented in CA#12 to the same CR revealed that there were no remaining issues which would prevent the closure of the PHAD White finding. |
| 05000334/FIN-2002003-02 | N/A | 2002Q1 | The licensee changed its emergency plan such that the PHADs were no longer considered a part of the siren notification system but were considered a supplemental part of the ANS. This change was determined to be a decrease in the effectiveness of the emergency plan. Decreases in the effectiveness of an emergency plan must receive NRC review and approval prior to implementation. The licensee entered this issue into its corrective action program (Condition Report 02- 02195) and will change the emergency plan back to the original wording. The implementation of a change which decreased the effectiveness of the emergency plan is being treated as a non-cited violation consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388). |
| 05000334/FIN-2002003-03 | N/A | 2002Q1 | The inspectors identified an issue related to the adequacy of corrective actions regarding the PHADs. In a 1998 audit, the licensee had identified that there was no procedure to formalize PHAD maintenance and testing. In a subsequent audit, it was determined that the 1998 audit finding had been closed without addressing the issue. The licensee then developed a procedure to address the initial issue, however, the audit and corrective actions were narrowly focused on adequacy of documentation of PHAD testing but did not consider overall PHAD operability. |
| 05000334/FIN-2006009-01 | Failure to Provide Adequate Corrective Actions to a PREVIOUSLY-IDENTIFIED Emergency Preparedness Exercise Weakness | 2006Q3 | The inspector identified an apparent violation for the licensees failure to provide adequate corrective actions to a previously-identified emergency preparedness exercise weakness. 10 CFR 50, Appendix E, Section IV.F.2.g, requires that any emergency preparedness weakness or deficiency that is identified shall be corrected. An apparent violation of that requirement was identified involving the licensees failure to adequately correct a performance deficiency in the area of Protective Action Recommendation development identified by the NRC in the May 2004 evaluated exercise. Specifically, in the 2006 exercise, the licensee dose assessment team did not adequately consider plant-specific situational information to develop the best dose projection estimate achievable at the time, which was an apparent repeat of a problem exhibited in the 2004 exercise. The licensees 2006 performance regarding the development of a dose projection without a sound technical basis demonstrated that the licensee had implemented ineffective corrective actions for the 2004 inspection finding. This finding is greater than minor because it is associated with the Emergency Response Organization Performance attribute and affected the objective of the Emergency Preparedness Cornerstone to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The Emergency Preparedness SDP, Section 5.3, Failure to Correct Drill or Exercise Weaknesses, was used to evaluate the significance of this finding. Because the licensees corrective actions were not adequate and the weakness involved a Risk Significant Planning Standard area that is not covered by performance indicators (i.e., 10 CFR 50.47(b)(9)), a loss of planning standard function was assessed, resulting in a White finding. |
| 05000334/FIN-2007005-02 | Weld Overlays on Pressurizer Safety Nozzles Not Initially Qualified for P-1 Materials | 2007Q4 | The inspectors observed the following NDE activities and reviewed completed NDE inspection data records to evaluate compliance with the ASME Code Section V and Section XI requirements and to verify that the indications and defects (if present) were dispositioned in accordance with the ASME Code Section XI requirements. Manual Ultrasonic Examinations (UT) of pressurizer relief valve nozzle weld overlay and pressurizer spray nozzle weld overlay and completed UT examination data records Manual UT examination inspection data report record review of UT-07-1011 Pressurizer longitudinal weld, location RC 740 PZR and Reactor in-vessel automated UT examination report and indication report record review of UT-07- 1012 circumferential weld, location RC-062A-780 Dye penetrant testing (PT) examination record PT-07-1006 review of reinforcement plate to nozzle weld RH-E-1A-N-5 (inlet) & RH-E-1A-N-6 (outlet) PT and magnetic particle testing (MT) examination record review of five pressurizer nozzles prior to weld overlays Bare Metal Inspection (BMI) visual examination records Report No. BOP-VT-07- 031 and digital photographic review of reactor vessel lower head penetrations inspection, and Visual Testing (VT) examination records review of Reactor Vessel and Internals Report No. VT-07-1142, VT-07-1144, VT-07-1145, and VT-07-1146. The inspectors reviewed pressure boundary welds for Class 1 or 2 systems which were completed since the beginning of the previous 1R17 refueling outage to determine if the welding acceptance and pre-service examinations (e.g., VT, PT, and weld procedure qualification tests) were performed in accordance with ASME Code Sections III, V, IX, and XI requirements. During the current 1R18 refueling outage, FENOC mitigated the pressurizer nozzle Alloy 82/182/600 welds to prevent Primary Water Stress Corrosion Cracking (PWSCC) induced through wall cracking in the Reactor Coolant System (RCS) pressure boundary. Mitigation activities included weld overlays on three safety valve nozzles, spray nozzle, and relief valve nozzle on the pressurizer. The inspectors remotely observed automated welding activities associated with the structural weld overlays of the pressurizer dissimilar metal welds on ASME Class 1 pressurizer piping nozzles. The inspectors reviewed procedures and records associated with the welding activity and observed the weld overlay process and ensured that the correct welding variable settings were being employed. In addition, certifications of the NDE technicians performing the manual-driven, encoded phased array UT examinations, as well as ASME Welder Maintenance Logs of the individual contractors performing the weld overlay activities on the pressurizer nozzles were reviewed. On October 2, during installation of weld overlays on pressurizer safety nozzles, it was discovered that welding was being performed with a procedure that had not been qualified for the application, and therefore, did not meet ASME Construction Codes (ASME Section III f65 edition, Winter f66 addenda, ASME Section IX, latest edition) requirements. The PCI Energy Services welding procedure WPS 3-8/52-TB MCGTAWN638 used for the P-1 portion of layer 1 of the weld overlays on pressurizer safety nozzles A, B, and C was not qualified to ASME Section III and IX requirements for P- 1 materials (Condition Report 07-27664). The implemented welding procedure was qualified for P-3 material; therefore, the contractor proceeded to qualify the welding procedure. FENOC made the decision to proceed with the weld overlays at risk while the procedure qualification testing was in progress. The weld procedure was subsequently qualified for P-1 material and acceptable for use. This issue remains unresolved until NRC completes its final evaluation of FENOCfs assessment of the use of the referenced procedure for P-1 material. (URI 05000334/2007005-02, Procedure used for Weld Overlays on Pressurizer Safety Nozzles was not Initially qualified for P-1 Material). |
| 05000334/FIN-2007005-06 | Inadequate 10 CFR 50.59 Review Results in Condition Beyond Design Basis During Test | 2007Q4 | The inspectors identified that the licensee did not perform an adequate safety evaluation in accordance with 10 CFR 50.59 associated with changing the periodicity of IST testing of valves MOV-1SI-890 A&B in May 2006. The review did not identify that the change allowed operations of these valves in Operational Modes where operation was prohibited by TS. The change was approved and implemented and as a result, from May 2006 until July 2007, valves MOV-1SI-890 A&B were cycled nine times total. Upon discovery, the licensee entered this issue into their corrective action program as CR 07-23462, conducted a root cause analysis and an extent of condition review, and revised the LHSI surveillance procedures. The licensee also determined that this event was reportable and issued LER 05000334/2007-001. The performance deficiency and violation is that the licensee did not perform an adequate safety evaluation in accordance with 10 CRF 50.59, due to the fact that the evaluation failed to identify that a change would proceduralize an operation which was prohibited by TS. This change would have required prior approval from the NRC via Technical Specification Amendment, to allow this change. A 10 CFR 50.59 violation is considered to potentially impede or impact the regulatory process; therefore, Traditional Enforcement applies. Comparing this item to the examples in NUREG 1600 Supplement I, this finding is more than minor because NRC approval would have been required. The inspectors completed a Significance Determination Review using IMC 0609, Appendix A Significance Determination of Reactor Inspection Findings for At Power Situations. Using the Phase I Screening worksheet the finding was determined to be of very low safety significance (Green) since the finding did not represent an actual loss of safety function for greater than the Technical Specification allowed outage time. Therefore, the finding is similar to Item D.5 in NUREG 1600 Supplement I, Violations of 10 CFR 50.59 that result in conditions evaluated as having very low safety significance (i.e., green) by the SDP. This is an example of a Severity Level IV violation. There is no cross cutting aspect for this finding, because it was determined that this finding is not reflective of current licensee performance. |
| 05000334/FIN-2008002-01 | Incorrect Jumper Placement During Testing Renders Quench Spray Chemical Addition Inoperable | 2008Q1 | A self-revealing, Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XI Test Control, was identified for the failure to properly perform line starter testing for a Unit 2 safety-related battery exhaust fan (2HVZ-FN216B) in accordance with the written test procedure. The test procedure for the line starter establishes test conditions by installing jumpers into a process rack, RK-2SEC-PROC-B1. Due to misidentification, jumpers were installed into the incorrect process rack, RK-2SEC-PROC-B. This rendered the B train of Quench Spray Chemical Additive System inoperable. The licensee entered the deficiency into their corrective action program as Condition Report 08-37168. FENOC performed a root cause evaluation, evaluated appropriate human performance contributors, and initiated corrective actions to prevent recurrence. The finding is greater than minor because it affected the equipment performance attribute of the associated Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance, because there was no overall loss of system function due to system redundancy and the system would have been able to perform its required safety function for the applicable mission time during design basis events. The cause of this finding is related to the cross-cutting area of human performance, in that FENOC failed to utilize adequate self and peer checking during the identification of equipment and circuits specified in the test plan (H.4(a)) |
| 05000334/FIN-2008003-01 | Inadequate Maintenance Procedure Results in Unexpected Terry Turbine Speed Increase | 2008Q2 | A self-revealing NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified in that the licensee failed to incorporate sufficient assembly detail into the maintenance procedure for the governor linkage on the Turbine-Driven Auxiliary Feedwater (TDAFW) pump. The required gaps and tightening criteria for the reassembly of the governor valve linkage were not included in the overhaul procedure resulting in jam nuts loosening, allowing the valve stem to rotate. Rotation of the valve stem caused an uncontrolled change in position of the governor valve position. This resulted in an unanticipated speed increase of the TDAFW during the performance of surveillance test 1OST-24.4 Steam Turbine Driven Auxiliary Feed Pump Test (1FW-P-2). Corrective actions included a change to the maintenance procedure and the installation of spacer shims for the anti-rotation block. This finding was more than minor because it affected the equipment performance attribute of the associated Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter (IMC) 0609, Attachment 609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low risk significance. The cause of this finding is related to the cross-cutting area of human performance, in that FENOC did not maintain a complete, accurate, and up-to-date governor overhaul procedure in regards to actuator reassembly which resulted in speed control degradation to the TDAFW (H.2.(c)) |
| 05000334/FIN-2008003-02 | Deficient Control of Clearance Posting Interrupts Reactor Coolant Charging Path While Vessel Water Level Drained Below the Flange | 2008Q2 | \"A self-revealing finding was identified for failure to properly coordinate clearance activities associated with testing for penetration 2X-46 during reduced reactor coolant system (RCS) level. A decision to post a clearance to support penetration testing resulted in the isolation of the make-up flow charging path to the reactor coolant system, resulting in an unexpected reduction of reactor coolant vessel level that was identified and stabilized within the established band. The licensees immediate corrective actions were to stop work, perform system configuration verification, and re-evaluated in progress and planned activities for plant safety impact. Long-term corrective actions include a change in procedures to not allow this type of penetration test in this plant configuration. The finding is more than minor because it affects the configuration control attribute of the Initiating Events cornerstone and affects the shutdown equipment lineup needed for stable reactor vessel level control during reduced RCS level operations, a high risk evolution. The inspectors performed a Phase 1 SDP evaluation in accordance with IMC 0609, Appendix G, Attachment 1, Checklist 3, Pressurized Water Reactor Cold Shutdown and Refueling Operation with RCS Open and Refueling Cavity Level < 23. The inspectors reviewed station drawings and records of reactor vessel level indication during the event. The inspectors determined that although make-up flow was momentarily isolated, reactor vessel level was maintained, sufficient indication existed, and no actual loss of RCS inventory occurred. Therefore, a Phase 2 quantitative assessment was not required and the issue screened to Green (very low safety significance). The cause of this finding is related to the cross-cutting area of human performance, in that FENOC did not appropriately coordinate work activities for the existing plant conditions to ensure the operational impact on reactor vessel level while at a reduced water level was fully understood (H.3(b)) |
| 05000334/FIN-2008003-03 | Failure to Properly Implement Operating Procedure During Plant Startup | 2008Q2 | A Green self-revealing NCV of TS 5.4.1.(a) was identified in that the licensee failed to take appropriate action to trip the main turbine as specified in 2OM-52.4.A, Raising Power from 5% to Full Load Operation, Rev. 13 during an unexpected main turbine load increase that resulted in average reactor coolant temperature below the operational limit of 541F. The licensee has developed and implemented an operations department rapid improvement plan. This finding was more than minor because it can be reasonably viewed as a precursor to a significant event. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter (IMC) 0609, Attachment 609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low risk significance. The cause of this finding is related to the cross-cutting area of human performance, in that FENOC failed to properly communicate critical parameters and limitations for personnel to perform work safely in a timely manner (H.1.(c)) |
| 05000334/FIN-2008003-04 | Failure to Properly Implement Abnormal Operating Procedure During Plant Startup | 2008Q2 | A Green self-revealing NCV of TS 5.4.1.(a) was identified in that the licensee failed to properly enter and implement the appropriate abnormal operating procedure (AOP) for loss of main feedwater. The licensee has developed and implemented an operations department rapid improvement plan. This finding was more than minor because it can be reasonably viewed as a precursor to a significant event. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with Inspection Manual Chapter (IMC) 0609, Attachment 609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low risk significance. The cause of this finding is related to the cross-cutting area of human performance, in that FENOC failed to properly implement appropriate roles and authority for decision making during risk-significant decisions. (H.1.(a)) |
| 05000334/FIN-2008003-05 | Licensee-Identified Violation | 2008Q2 | \"10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, requires that, Measures shall be established to assure that special processes, including welding, heat treating and nondestructive testing, are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements. Contrary to 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, the licensee failed to assure that a special welding process was performed in accordance with the applicable ASME Code requirements, the PCI Energy Services welding procedure WPS 3-8/52-TB MCGTAW-N638 used for the P-1 portion of layer 1 of the weld overlays on pressurizer safety nozzles A, B, and C was not qualified to ASME Section III and IX requirements for P-1 materials. Failure to comply with ASME Code welding requirements could result in flaws within reactor coolant system piping welds. This issue was documented in the licensees corrective action program as Condition Report 07-27664. The safety significance of this issue was considered very low, since there were no recordable indications in the weld overlay, no adverse consequences were identified, the procedure deficiency was found prior to returning the component or system to service, and the welding process was subsequently qualified. Failure to perform welding of Class 1 welds in compliance with the applicable American Society of Mechanical Engineers (ASME) Construction Codes (ASME Section III 65 edition, Winter 66 addenda, ASME Section IX, latest edition) requirements is considered a licensee-identified violation (Green), Non-Cited Violation of 10CFR Part 50, Appendix B, Criterion IX, Control of Special Processes \ |
| 05000334/FIN-2008004-01 | Procedure User Errors Result in Loss of an Electrical Bus. | 2008Q3 | A self-revealing NCV of TS 5.4.1.(a), Procedures, was identified in that FENOC failed to properly implement procedures and required actions in planning, tagging, and electrical system operation. A series of procedural use errors in control of maintenance, equipment control and electrical system operation resulted in the inadvertent loss of the 1G 4160VAC (4kV) electrical bus. The licensee remediated the operating crew and communicated station expectations regarding organizational interfaces and procedural compliance. This was also communicated to all station crews, maintenance, and construction services departments. This finding is more than minor because it is similar to Inspection Manual Chapter (IMC) 0612, Appendix E, example 3b, since the procedural use errors resulted in the loss of the 1G Bus. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with IMC 609, Attachment 609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low risk significance. The cause of this finding is related to the cross-cutting area of human performance, in that FENOCs failed to follow station procedures resulting in a loss of the 1G bus (H.4.(b)). (Section 4OA2.1) |
| 05000334/FIN-2008004-02 | Licensee-Identified Violation | 2008Q3 | 10 CFR 55.21, Medical Examination, requires in part that, A licensee shall have a medical examination by a physician every two years. This period expires at the end of the calendar month of the two year anniversary of the previous physical. Contrary to 10 CFR 55.21, Medical Examination, the licensee identified that two licensed reactor operators had expired licensed operator physicals. One licensed reactor operator had performed licensed duties for thirty two (32) hours after the physical had elapsed. The other licensed operator had not performed licensed duties with an expired physical. The licensee took immediate corrective actions to have the operators examined by a physician and there were no adverse changes in the operators physical conditions since the last physical. Based upon this, the violation was of very low safety significance. The licensee entered this issue into their corrective action program as CRs 08-45075 and 08-45291. This is a licenseeidentified violation (Green), NCV of 10 CFR 55.21, Medical Examination. |
| 05000334/FIN-2008008-01 | Inadequate Corrective Action for Potential Blockage of AFW Pump Lube Oil Cooling System Orifices When Supplied by Rw/Sw | 2008Q4 | The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, in that FENOC did not take adequate corrective action following the identification of a condition adverse to quality. Specifically, in 2004, 2005 and 2006, FENOC identified that if the Unit 1 river water (RW) system or the Unit 2 service water (SW) system was aligned to the suction of the auxiliary feedwater (AFW) pumps it could result in blockage of cooling water flow for the pumps, but did not take actions to correct the deficiency. FENOC entered the issue into their corrective action program to correct the non-conformance. In addition, FENOC developed Operations Department standing orders to limit the use of TS action statement 3.7.6.a which credited the use of the lineup, and formalized compensatory actions to address an Appendix R compliance deficiency. The finding was more than minor because there was reasonable doubt as to the operability of the AFW system when supplied from RW or SW systems. In addition, the finding was associated with the design control attribute of the Mitigating Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding represented a potential loss of safety function, the team conducted Significance Determination Process (SDP) Phase 2 and Phase 3 analyses which determined the finding was of very low safety significance (Green). Finally, the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program Component, because FENOC did not adequately evaluate this condition adverse to quality, including classifying, prioritizing, and evaluating for operability when it was identified in February 2004, and again in March 2005 and June 2006. (IMC 0305, Aspect P.1(c)). (Section 1R21.2.1.1 |
| 05000334/FIN-2008502-01 | Licensee-Identified Violation | 2008Q3 | 10 CFR 50.54(q) requires licensees to follow and maintain in effect an emergency plan which meets the standards of 10 CFR 50.47(b). The Beaver Valley Power Station Emergency Plan states that emergency action level values are based upon criteria established under NUMARC/NESP-007, Methodology for Development of Emergency Action Levels. In March and October 2006 FENOC made changes to the Unit 1 and Unit 2 emergency action levels (EALs), respectively, as a result of the units extended power up-rates. The licensee used calculation package 10080- UR(B)-507, Containment High Range Radiation Area Monitor Readings Due to LOCA With Various Source Terms Addresses Alternative Source Terms and Power Up-rate, to revise area radiation monitor thresholds in the EAL fission product barrier matrix. In making the changes, the licensee committed two separate errors. The first error was that the values for the containment radiation monitors listed under the Fuel Cladding Barrier (EAL 1.1.6) were taken from the wrong table in the calculation package, and this affected both Units. The second error was an incorrect mathematical conversion that involved the threshold for Significant Radioactivity in Containment (EAL 1.3.5), which only affected Unit 2. Upon discovery of this error, in April, 2008, the licensee took immediate action to correct the related EAL radiation values and issued CR 08-38146, which initiated a root cause evaluation and a technical evaluation to understand what redundant EAL thresholds may have been exceeded before the affected EAL 1.1.6 and 1.3.5 thresholds. The inspector determined that the errors associated with these EAL parameters were mathematical in nature and were of very low safety significance because they would not have delayed the declaration of any event, due to redundant EALs. |
| 05000334/FIN-2009002-01 | Licensee-Identified Violation | 2009Q1 | Beaver Valley Unit 1 and Unit 2 TS 3.7.10 requires, in part, that an adequate Control Room Envelope (CRE) be maintained or restored within 90 days. This protects the CRE during postulated accident and hazardous conditions. Contrary to TS 3.7.10, the licensee determined that an inadequate CRE existed, due in part to a degraded (damper corrosion) normal intake damper, which is postulated to have existed for longer than the TS allowed time. The separate Control Room Emergency Ventilation System (CREVS) was not affected. The licensee identified the excessive inleakage condition during a surveillance test. The licensee had failed to to identify this component as a CRE boundary and perform routine inspection and maintenance. Upon finding the excessive in-leakage, the licensee implemented compensatory actions to mitigate the possibility of in-leakage to the control room and completed repairs to the affected dampers and seals. The licensee entered this issue into their corrective action program as CR 08-49260 and reviewed their CRE maintenance program. The inspectors determined that the failure to maintain an adequate CRE is a violation of TS 3.7.10 identified by the licensee that affects the containment barrier cornerstone. The violation is considered of very low safety significance since it represents a degradation of the radiological barrier of the control room. This is considered a licensee-identified violation (Green), NCV of Technical Specification 3.7.10 |
| 05000334/FIN-2009002-02 | Licensee-Identified Violation | 2009Q1 | Beaver Valley Unit 2 TS 3.6.2 action A.2 requires that the operable door for a primary containment airlock is to be locked closed within 24 hours in the event that a primary containment airlock door becomes inoperable. This action is to prohibit inadvertent use of the operable airlock door to maintain primary containment integrity with an inoperable airlock door. Contrary to TS 3.6.2 action A.2, the licensee identified that the airlock to atmosphere equalizing valve 2PHS-111 was open and out of its normal position from September 21 to September 25. This rendered the outer airlock door inoperable for the same period of time. The licensee failed to lock the inner airlock door as required within 24 hours. Upon finding 2PHS-211 open during the performance of 2OST-47.1 Containment Air Lock Door Test, the licensee took the immediate action of closing 2PHS-111 to restore the outer airlock door to operable status. The inner airlock door remained closed for the entire time and satisfactorily maintained the primary containment penetration operable. The licensee entered this issue into their corrective action program as CR 08-46883. Subsequently, the licensee conducted a root cause evaluation and has determined that the level of procedure use to operate the airlock doors should be Step-by-Step procedures, and implemented procedure changes. The violation is considered of very low safety significance because it affects the Containment Barrier Cornerstone and is a Type B finding that initially screens out of Manual Chapter 0609, Appendix H, Table 4.1 as Green. Because the licensee discovered 2PHS-111 open during the performance of surveillance test 2OST-47.1, this is considered a licensee-identified violation (Green), NCV of Technical Specification 3.6.2 |
| 05000334/FIN-2009003-01 | Inadequate Post-Maintenance Testing Specified for Safety-Related River Water Check Valve | 2009Q2 | A non-cited violation (NCV) of 10CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings was identified for failure to specify and perform an adequate post-maintenance test (PMT) after replacing a safety-related river water check-valve. Specifically, the PMT under work order 200233562 was not adequate to verify the proper function of the valve 1RW-57 prior to its return to service. The PMT was subsequently performed successfully. This issue was entered into the licensees corrective action program as condition report 09-59866. The failure to specify and perform an adequate PMT after replacing a safety-related river water check-valve was a performance deficiency. The finding was more than minor in accordance with IMC 0612, Appendix B (Section 1-3), Issue Screening, because the failure to specify and perform an adequate PMT is associated with the procedure quality performance attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding has a cross-cutting aspect in the area of human performance associated with resources because the licensee did not have complete, accurate, and up-to-date maintenance work procedures (IMC 0305 Aspect: H.2(c) |
| 05000334/FIN-2009003-02 | Continuously Submerged Cables Design Deficiency | 2009Q2 | The inspectors identified a non-cited violation (NCV) of 10CFR Part 50, Appendix B, Criterion III, Design Control, in that FENOC failed to maintain safety-related cables in an environment for which they were designed. Since NRC Information Notice 2002-12 was issued, FENOC has had several opportunities to trend as-found data, implement effective maintenance programs, and identify and thoroughly evaluate long-term adverse conditions for underground safety-related cables exposed to continuous submerged environments. Cables affected include those for Unit 1 river water and Unit 2 service water. The issue was entered into the licensees corrective action program (CR 09-60496) to initiate a review of the current manhole and cable monitoring programs, and to initiate long-term corrective actions. Failure to maintain safety related cables in an environment for which they were designed is considered a performance deficiency. The finding was more than minor in accordance with IMC 0612, Appendix B (Section 1-3), Issue Screening, because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern. Specifically, the deficiency did not result in the present loss of operability or functionality and did not represent a risk significant external event such as flooding. The issue was entered into the licensees corrective action program (CR 09-60496) to initiate a review of the current manhole and cable monitoring programs, and to initiate long-term corrective actions. The performance deficiency had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not thoroughly evaluate problems such as resolutions, address causes, and evaluate the effectiveness of corrective actions (IMC 0305 Aspect: P.1 (c) |
| 05000334/FIN-2009003-03 | Licensee-Identified Violation | 2009Q2 | Technical Specification 3.6.1, Containment, requires that containment operability be maintained in Mode 1, restored within one hour, or the reactor be shutdown to Mode 3 within six hours. Contrary to this requirement, FENOC failed to maintain containment operability or restore containment operability in the allowed time. Specifically, FENOC did not ensure containment isolation valves MOV-1RS-155B and MOV-1RS-156B were closed and de-energized prior to opening the 1RS-P-2B pump casing drain valve. The issue was entered into FENOCs corrective action program as CR 09-56250. The finding was more than minor because it is associated with the configuration control attribute of the barrier integrity cornerstone and affects the cornerstone objective of ensuring containment boundary preservation under postulated design-basis accident scenarios. The inspectors determined that the finding was of very low safety significance (Green), based on IMC 0609, Appendix H, Table 4.1 because this is a Type B finding and the affected pipe size is less than 2 inches in diameter |
| 05000334/FIN-2009005-01 | Failure to Properly Verify Valve Line-up Results in an Overpressurization of the Isolated CIS\" Reactor Coolant Loop. | 2009Q4 | A self-revealing NCV of TS 5.4.1, Procedures, was identified in that operators failed to properly align and check the position of the B reactor coolant system (RCS) loop bypass valve (2RCS*45). as required by procedure. This deficiency caused an incorrect lineup of the required vent path and resulted in the over-pressurization of the isolated B RCS loop while filling. The estimated pressure exceeded the pressurel temperature limit for an isolated RCS loop on November 11. The inspectors determined that the finding was not similar to the examples for minor deficiencies contained in IMC 0612, Appendix E, Examples of Minor Issues, The finding was more than minor because if left uncorrected could have the. potential to lead to a more significant safety concern. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for ill)pacting NRCs regUlatory function, and was not the result of any willful violation of NRC requirements. The inspectors performed a Phase 1 SOP evaluation in accordance with fMC 0609, Appendix G, Attachment 1, CHECKLIST 4 PWR Refueling Operation: RCS level> 23 OR PWR Shutdown Operation with Time to Boil> 2 hours And Inventory in the Pressurizer. Because the loop was isolated from the reactor vessel and pressurizer, the required reactor coolant inventory and the decay heat removal system were not affected. There were no conditions indicating a loss of control as listed in Appendix G, Table 1 Losses of Control. Therefore. a Phase 2 quantitative assessment was not required and the issue screened to Green (very low safety significance). Because this finding is of very low safety significance and has been entered into FENOCs corrective action program, the violation is being treated as a non-cited violation. The cause of this find(ng is related to the cross~cutting area of human performance, work practices, in that FENOCs failed to follow station procedures resulting in an overpressurization of the isolated B RCS loop. (H.4.(b)). |
| 05000334/FIN-2009005-02 | Inadequate RHS Shutdown Procedure Results in Unusual Event Declaration. | 2009Q4 | A self-revealing NCV of TS 5.4.1, Procedures, was identified in that procedures for securing Residual Heat Removal System (RHS) were not adequately maintained and did not contain relevant operating restrictions resulting in the inadvertent lifting of the A RHS pump suction relief(2RHS-RV721A) during normal operation, excessive identified leakage of reactor coolant to the Pressurizer Relief Tank, and a declaration of an Unusual Event. The inspectors determined that the finding was not similar to the examples for minor deficiencies contained in IMC 0612, Appendix E, Examples of Minor Issues. The finding was more than minor because if left uncorrected could have the potential to lead to a more significant safety concern. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRCs regulatory function, and was not the result of any willful violation of NRC requirements. The inspectors performed a Phase 1 SOP evaluation in accordance with IMC 0609, Appendix G. There were no conditions indicating a loss of control as listed in Table 1 .Losses of ControL Attachment 1, Checklist 1 PWR Hot Shutdown Operation: Time to Core Boiling <2 Hours guidelines were used to evaluate the event. All mitigating capabilities were available, therefore a Phase 2 quantitative assessment was not required. The issue screens to Green (very low safety significance). Because this finding is of very low safety significance and has been entered into FENOCs corrective action program (CR 09-68214), the violation is being treated as a non-cited violation. The cause of this finding is related to the cross-cutting area of human performance, resources, in that procedures for RHS system shutdown were not complete and up to date. (H.2(b)) |
| 05000334/FIN-2009007-01 | Failure to Evaluate the Impact of Spurious Volume Control Tank Outlet Valves Closure During Alternative Shutdown | 2009Q3 | Beaver Valley Unit 2 License Condition 2.F requires in part that FENOC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report (FSAR) through Amendment No. 17, and submittals dated May 18, May 20, May 21, June 24, and July 6, 1987. BVPS Unit 2 FSAR Rev. 17 Section 9.5.1.1 states that the fire protection system is designed using the guidance of BTP CMEB 9.5-1, Rev. 2. BTP CMEB 9.5-1, Rev. 2, Section C.5.c.(7) requires that alternative safe shutdown equipment and systems for each fire area should be known to be isolated from associated circuits in the fire area so that hot shorts, open circuits, or shorts to ground in the associated circuits will not prevent operation of the safe shutdown equipment. Contrary to the above, on August 4, 2009, the NRC identified that FENOC did not meet this requirement and failed to protect the 21A charging pump from a single circuit failure resulting in the spurious closure of the VCT outlet valves. FENOC is in transition to NFPA 805 and therefore the NRC-identified violation was evaluated in accordance with the criteria established by Section A of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR Part 50.48) for a licensee in NFPA 805 transition. Specifically, although the NRC identified the violation, it is likely that FENOC would have identified and corrected this issue as part of the transition to NFPA 805, FENOC entered the issue into the corrective action program, FENOC implemented compensatory measures in a reasonable time commensurate with the risk significance, the issue was not likely to have been previously identified by routine FENOC efforts, and the violation was not willful. Because all the criteria were met, the NRC is exercising enforcement discretion for this issue. |
| 05000334/FIN-2009007-02 | Failure to Evaluate the Affects of a Spurious Safety Injection Signal During Alternative Shutdown | 2009Q3 | Beaver Valley Unit 2 License Condition 2.F requires in part that FENOC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report (FSAR) through Amendment No. 17, and submittals dated May 18, May 20, May 21, June 24, and July 6, 1987. BVPS Unit 2 FSAR Rev. 17 Section 9.5.1.1 states that the fire protection system is designed using the guidance of BTP CMEB 9.5-1, Rev. 2. BTP CMEB 9.5-1, Rev. 2, Section C.5.c.(1) requires that during postfire shutdown, the fission product boundary integrity shall not be affected; i.e., there shall be no rupture of any primary coolant boundary. Contrary to the above, on July 23, 2009, the NRC identified that FENOC did not meet this requirement and failed to ensure that the RCS SRVs would remain closed if a single spurious operation of the SI signal were to occur during a postfire shutdown. FENOC is in transition to NFPA 805 and therefore the NRC-identified violation was evaluated in accordance with the criteria established by Section A of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR Part 50.48) for a licensee in NFPA 805 transition. Specifically, although the NRC identified the violation, it is likely that FENOC would have identified and corrected this issue as part of the transition to NFPA 805, FENOC entered the issue into the corrective action program, FENOC implemented compensatory measures in a reasonable time commensurate with the risk significance, the issue was not likely to have been previously identified by routine FENOC efforts, and the violation was not willful. Because all the criteria were met, the NRC is exercising enforcement discretion for this issue. |
| 05000334/FIN-2009007-03 | Failure to Ensure RCPs were Tripped and Remain Tripped Prior to Securing Reactor Coolant Pump Seal Cooling | 2009Q3 | Beaver Valley Unit 1 License Condition 2.C.5 requires that FENOC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report. BVPS Unit 1 UFSAR Rev. 24 Section 9.10.1 states that the fire protection plan that satisfies General Design Criterion 3 of Appendix A to 10 CFR 50 is described in BVPS Administrative Procedures. 1/2 ADM-1900, Fire Protection Program, Rev. 19 step 7.17.4 requires operating procedures be maintained to implement the actions required to achieve safe shutdown. Contrary to the above, on July 29, 2009, the NRC identified that FENOC did not meet this requirement and failed to maintain safe shutdown procedure 1OM-56B.4.I, Safe Shutdown Following a Serious Fire in the Service Building, Rev. 11, and ensure that the RCPs were tripped and remain tripped prior to securing RCP seal cooling. FENOC is in transition to NFPA 805 and therefore the NRC-identified violation was evaluated in accordance with the criteria established by Section A of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR Part 50.48) for a licensee in NFPA 805 transition. Specifically, although the NRC identified the violation, it is likely that FENOC would have identified and corrected this issue as part of the transition to NFPA 805, FENOC entered the issue into the corrective action program, FENOC implemented compensatory measures in a reasonable time commensurate with the risk significance, the issue was not likely to have been previously identified by routine FENOC efforts, and the violation was not willful. Because all the criteria were met, the NRC is exercising enforcement discretion for this issue. |
| 05000334/FIN-2009008-01 | Containment Isolation Valve System 10 CFR 50.65 (a)(2) Performance Demonstration Not Met. | 2009Q3 | The inspectors identified an NCV of. very low safety significance (Green) of 10 CFR 50.65(a)(2), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, due to FENOC personnels failure to demonstrate that the 10 CFR 50.65(a)(2) performance of the containment isolation valve limit switches was effectively controlled through the performance of appropriate preventive maintenance. Specifically, as evidenced by repeat dual position indications of containment isolation valves in the control room between 2007 and 2009 resulting in 21 unplanned entries into Technical Specification 3.6.3, the containment isolation valve system 10 CFR 50.65(a)(2) performance demonstration was no longer justified in accordance with Maintenance Rule implementing procedure guidance. This should have resulted in placement of the containment isolation valve system in 10 CFR 50.65(a)(1) for goal setting and monitoring. FENOC entered this issue into the CAP (CR 09-64040). The inspectors determined the finding was more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the finding did not involve a design or qualification deficiency resulting in loss of operability or functionality, did not result in a loss of system safety function, and did not screen as potentially risk significant due to external initiating events. the inspectors determined that this finding had a cross-cutting aspect in the Corrective Action Program component of the Problem Identification and Resolution cross-cutting area because FENOC did not take appropriate corrective actions to address safety issues and adverse trends associated with faulty containment isolation valve limit switches in a timely manner, commensurate with their safety significance and complexity (P.1 (d)) |
| 05000334/FIN-2010002-01 | Human Performance Error Results in Disabling a Control Room Annunciator | 2010Q1 | A self-revealing finding was identified for FENOCs failure to properly implement a station procedure. Specifically, work order instructions were not properly followed, as specified in NOP-WM-4006, Conduct of Maintenance, causing leads to be inadvertently lifted for an alarm to the main control room control board. This annunciator is used by operators in the Loss of Main Feedwater Abnormal Operating Procedure. The leads were reconnected and this issue was entered into the licensees corrective action program as CR 10-72654. The finding is more than minor because it is similar to example 2.f in IMC 0612, Appendix E. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRCs regulatory function, and was not the result of any willful violation of NRC requirements. In accordance with IMC 0609.04 (Table 4a), Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance. The cause of this finding relates to the cross-cutting aspect of Human Performance, Work Practices, in that FENOC personnel did not follow procedures, resulting in a control room annunciators leads being inadvertently lifted. (HA.(b) |
| 05000334/FIN-2010004-01 | Inadequate Maintenance Procedure Results in Auto-Disassembly of EDG Intake Damper | 2010Q3 | A Green, self-revealing finding (FIN) was identified in that an inadequate procedure resulted in a failure to adequately retain a 1-1 Emergency Diesel Generator (EDG) room damper after louver adjustment. Specifically, the adjustment of the 1-1 EDG upper damper (1VS-D-22-2A) in April 2010 led to retention hardware not being sufficiently secure to prevent damper failure and resulted in the linkage failing to open the upper dampers. This was selfrevealing during a crew investigation for a 1-1 EDG alarm on September 5,2010. This issue was entered into the licensee\'s corrective action program under CR 10-82257. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affects the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04 (Table 4a), Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green). The cause of this finding relates to the cross-cutting aspect of Human Performance, Resources, in that FENOC did not provide complete procedures to conduct the damper adjustment and retention. (H.2.(c)) |
| 05000334/FIN-2010005-01 | Premature Lifting of ECCS Relief Valve RV-1 SI-8458 | 2010Q4 | An unresolved item (URI) was identified because additional information regarding the existence of adverse nozzle loads affecting, and the as-found testing and inspection results of, the set point of Unit 1 relief valve RV-1-SI-8456 is required to determine whether a performance deficiency existed which contributed to its premature lifting. The inspectors will review the additional information after the relief valve is removed and the valve and associated pipe alignment is inspected. As reported by FENOC on November 16, 2010 (Event Notification 46421), and LER 05-334/2010-003-00, Unit 1 Low Head Safety Injection (LHSI) system relief valve RV- 1SI-8456 unexpectedly lifted during planned testing of the \\\'A\\\' LHSI pump. The relief valve properly reseated and cycled open only during pump operation. The condition revealed itself during surveillance testing by an unexpected sump alarm and lowering of the refueling water storage tank level. FENOC evaluated the impact to the LHSI safety function and the control room envelope. FENOC also determined that the relief valve could be gagged closed and still maintain LHSI system protection and testing capability with the remaining two relief valves and additional compensatory measures. This relief valve had been installed during the most recent refueling outage and post-maintenance tested satisfactorily. FENOC entered the issue into their CAP as CR 10-85863 and conducted a root cause analysis and concluded the most probable cause was nozzle loading effects from possible LHSI piping misalignment. FENOC plans to remove the valve and inspect it and the piping alignment at the earliest acceptable plant condition to continue its investigation. The inspectors will review the additional information of the as-found condition of the relief valve and the associated LHSI piping alignment to determine if a performance deficiency exists. |
| 05000334/FIN-2010005-02 | Failure to Follow Procedure Results in Main Feedwater Piping Pressurization | 2010Q4 | A self-revealing non-cited violation (NCV) was identified in that a chemical addition pump (1WT-P-15B) was misaligned to an isolated main feed water header, and upon starting caused an unexpected pressure transient. which affected the \'B\' Fast Acting Main Feedwater Isolation Valve (HYV-1FW-100B) (MFIV). Specifically, the main feed water piping was inadvertently isolated and Pressurized beyond its normal operating pressure, causing significant packing leakage of the \'B\' MFIV. This issue was entered into the licensee\'s corrective action program under CR 10-84891. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The inspectors determined that the finding was not similar to the examples for minor deficiencies contained in IMC 0612, Appendix E, Examples of Minor Issues. The finding was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed a Phase 1 SOP evaluation in accordance with IMC 0609, Appendix G, Attachment 1, Checklist 3 PWR Cold Shutdown and Refueling Operation RCS Open and Refueling Cavity Level <23\' OR RCS Closed and No Inventory in Pressurizer with Time to Boiling <2 hours. There was no loss of control, and all mitigating capabilities were available, therefore a Phase 2 quantitative assessment was not required and the issue screened to Green (very low safety significance). The cuase of this finding relates to the cross-cutting aspect of Human Performance, Work Practices, in that FENOC did not utilize human error prevention techniques, pre-job brief and peer checking, to prevent the misalignment of the chemical addition pump. |
| 05000334/FIN-2010005-03 | Maximum Differential Temperature Exceeded for Spray Nozzle during Pressurizer Heatup | 2010Q4 | A self-revealing non-cited violation (NCV) of TS 5.4.1, Procedures, was identified in that the shift technical advisor\'s (STA) failure to follow procedure resulted in the maximum differential temperature being exceed on the spray nozzle during pressurizer heat up. Specifically, the STA failed to notify the shift manager promptly when it became apparent that the maximum differential temperature of the spray nozzle trend was degrading and its limit subsequently exceeded. This issue was entered into the licensee\'s corrective action program under CR 10-85021. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The inspectors determined that the finding was not similar to the examples for minor deficiencies contained in IMC 0612, Appendix E. Examples of Minor Issues. The finding was more than minor because if left uncorrected, had the potential to lead to a more significant safety concern. The inspectors performed a Phase 1 SOP evaluation in accordance with IMC 0609, Appendix G, Attachment 1, Checklist 4 PWR Refueling Operation: RCS level> 23\' or PWR Shutdown Operation with Time to Boil> 2 hours And Inventory in the Pressurizer. There was no loss of control. aU mitigating capabilities were available, therefore a Phase 2 quantitative assessment was not required and the issue screened to Green (very low safety significance). The cause of this NCV relates to the cross-cutting aspect of Human Performance, Resources, in that FENOC personnel were not adequately trained to recognize the indications being monitored, resulting in the pressurizer spray nozzle maximum differential temperature being exceeded. |
| 05000334/FIN-2010005-04 | Licensee-Identified Violation | 2010Q4 | Violation of TS 5.4.1, \'Procedures\' for not properly calibrating the Unit 1 power range nuclear instrument channel N41 and returning it to an operable status following I&C subsequent to the post-outage maintenance on November 4. The instrument was not calibrated properly due to mis-direction on which portion of the procedure to perform. The reactivity senior operator stationed at-the-controls quickly recognized the degraded instrument and alerted the crew. The licensee documented this issue in CR 10-85405. The violation is not greater than green because only one channel was affected for a short period of time. |
| 05000334/FIN-2010005-05 | Licensee-Identified Violation | 2010Q4 | Violation of TS 5.4.1, \'Procedures\' for not properly conducting and documenting maintenance on the Backup Indicating Panel, a remote panel required for safe shutdown of Unit 1. Two thermocouple input wires were discovered disconnected internal to the panel on October 31 during a refueling surveillance test from a previous maintenance activity. The licensee documented this issue in CR 10-85168. The violation is not greater than green because alternate indications were available and simple repairs could be made in a timely manner. |
| 05000334/FIN-2010005-06 | 270 Degree Circumferential Flaw Found on Residual Heat Removal System Drain Valve Socket Weld | 2010Q4 | On October 2,2010, while Beaver Valley, Unit 1 was shut down in Mode 5 (cold shutdown) for a refueling outage, an active boric acid leak was identified on a drain valve (1 RH-200) located on the common suction piping for the Residual Heat Removal (RHR) system. ApprOXimately five hours after the initial identification of the active leak and entry into Mode 5, a nondestructive examination (NDE) was performed on the valve and its associated piping. A circumferential crack of 270 degrees in length with water seeping from the toe of the weld was discovered and both trains of RHR were declared inoperable. Since the adverse condition was identified prior to the completion of the NDE and no action was immediately taken as required by TS 3.4.7 for no required RHR trains operable while in Mode 5 operation, the licensee inadvertently entered a condition prohibited by TS 3.4.7. When the licensee determined that both trains of RHR were inoperable, they evaluated the issue for reportability and appropriately issued LER 50-334/2010-002, 270 Degree Circumferential Flaw Found on Residual Heat Removal System Drain Valve Socket Weld, dated November 29,2010. This LER reported that Beaver Valley, Unit 1 had been in a condition which was prohibited by TS 3.4.7.C, which requires immediate action if both trains of RHR are inoperable while in Mode 5 operations. The issue is considered within the traditional enforcement process because there was no performance deficiency identified and Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening directs disposition of this issue in accordance with the Enforcement Policy. The inspectors ysed the Enforcement Policy, Section 6.1 - Reactor Operations, to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor and best characterized as Severity Level IV (very low safety significance) because it is similar to Enforcement Policy Section 6.1, example d.1. Additionally, the inspectors assessed the risk associated with the issue by using (MC 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors screened the issue, and evaluated it using Checklist 2 of IMC 0609, Appendix G, Attachment 1. Throug out the duration of the event, the secondary side water level of at least two steam generators sufficient for decay heat removal (including necessary support systems) were available. As a result, this issue would screen as very low safety significance (Green). Because it has been determined that it was not reasonable for the licensee to be able to foresee and prevent the weld crack, or to have made the RHR inoperability decision at an earlier time, and as such no performance deficiency exists, the NRC has decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of TS (EA-11-004). Further, because licensee actions did not contribute to this violation, it will not be considered in the assessment process or the NRC's Action Matrix. |
| 05000334/FIN-2010201-01 | Security | 2010Q1 | |
| 05000334/FIN-2010403-01 | Security | 2010Q3 | |
| 05000334/FIN-2010502-01 | Licensee-Identified Violation | 2010Q2 | 10 CFR 50.54(q) requires that a nuclear power reactor licensee shall follow and maintain in effect emergency plans that meet the standards in 10 CFR 50.47(b). Contrary to this requirement, FENOC failed to maintain their described method of notifying State and local organizations of emergency events. Specifically, for the time period between February 8 and February 16, 2010, the Beaver Valley control room staff did not possess the newly-issued code word for verifying notification communications between the site and local officials. The control room staffs inability to use the proper code word would have unnecessarily delayed the required notification of those local officials, as required by 10 CFR 50.47(b)(5). Once discovered by FENOC on February 16, 2010, the condition was immediately corrected. FENOC further performed a Cause Analysis in conjunction with CR 10-71680 and initiated programmatic corrective actions to assure . the control room is properly informed when a new code word is assigned. The finding was more than minor because it is associated with the emergency preparedness cornerstone and affected the cornerstone objective of ensuring proper offsite notification of an emergency event at the Beaver Valley Nuclear Power Station. The inspectors determined that the finding was of very low safety significance (Green), based on IMC 0609, Appendix B, because, while not having the proper code word could have delayed theoffsite notification, alternate means of establishing offsite notifications were available and would have accomplished the notification objective. |
| 05000334/FIN-2011002-01 | lnadvertent Auxiliary Feedwater Start during Steam Generator Water Level Instrument Adjustments | 2011Q1 | A Green, self-revealing non-cited violation (NCV) of TS 5.4.1, Procedures, was identified in that technicians inadvertently caused an auxiliary feedwater actuation when the plant was shutdown for refueling. Specifically, the procedure used to inject simulated steam generator water signals was inadequate, which resulted in the technicians erroneously inserting two-of-three (213) Low-Low level signals to the SGs and causing actuation of auxiliary feedwater. This issue was entered into the licensee\'s corrective action program under CR 1 1-90528. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The finding is more than minor because it is similar to example 4.b in IMC 0612, Appendix E. The inspectors performed a Phase 1 SDP evaluation in accordance with IMC 0609, Appendix G, Attachment 1, Checklist 2 PWR Cold Shutdown Operation: RCS Closed and SGs Available for DHR (Loops Filled and Inventory in Pressurizer) Time to Boiling Less than 2 Hours. There was no loss of control, and all mitigating capabilities were available, therefore a Phase 2 quantitative assessment was not required and the issue screened to Green (very low safety significance). The cause of this NCV relates to the cross-cutting aspect of Human Performance, Work Control, in that FENOC personnel did not appropriately coordinate work activities by incorporating actions to address the need to communicate, coordinate and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance. |
| 05000334/FIN-2011002-02 | lnadequate Spray Additive System Sampling Procedures | 2011Q1 | A Green, self-revealing non-cited violation (NCV) of TS 5.4.1, Procedures, was identified in that chemistry procedures failed to provide adequate detail to ensure timely completion of TS required sampling of the spray additive system. Specifically, FENOC failed to complete timely sampling and analysis of the chemical addition tank, resulting in reasonable doubt of the operability of the spray additive system tor 13 days. The issue was entered into the licensee\'s corrective action program under CR 10-87438. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The finding is more than minor because it is similar to example 3.j in IMC 0612, Appendix E and it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04 (Table 4a), Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency which resulted in a loss of safety function. The cause of this NCV relates to the cross-cutting aspect of Problem identification and Resolution, Corrective Action Program, in that FENOC personnel did not implement a corrective action program with a low threshold for identifying issues. FENOC did not identify the issue completely, accurately and in a timely manner commensurate with its safety significance. |
| 05000334/FIN-2011003-01 | Failure to Maintain Recirculation Spray Heat Exchangers in Chemical Wet Layup | 2011Q2 | A Green, self-revealing non-cited violation (NCV) of TS 5.4.1, Procedures, was identified in that the Unit 2 Recirculation Spray System (RSS) Heat Exchangers (HXs) were not maintained in chemical wet layup, contrary to station procedures and industry guidance. Specifically, FENOC failed to maintain place corrosion inhibitors in the RSS HXs, resulting in significant HX corrosion, which led to degraded flow through the B RSS HX during a service water full flow test. This issue was entered into the licensee\'s corrective action program under CR 11-90430. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The inspectors determined that the finding was not similar to the examples for minor deficiencies contained in IMC 0612, Appendix E, Examples of Minor Issues. The finding is more than minor because it affects the Mitigating Systems and Barrier Integrity cornerstones. The finding is associated with the equipment performance attribute of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is also associated with the Structure Systems and Components (SSC) and barrier performance attribute of the Barrier Integrity cornerstone 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. In accordance with IMC 0609.04 (Table 4a), Phase 1 -Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because the finding did not result in a loss of operability, nor was it a degradation of a radiological barrier, control room barrier, hydrogen igniter, or an open pathway. The cause of this NCV relates to the cross-cutting aspect of Human Performance, Work Control, in that FENOC personnel did not coordinate work activities consistent with nuclear safety. Specifically, FENOC did not coordinate work activities to complete the chemical wet layup condition to support long-term equipment reliability of the Unit 2 RSS HXs. |
| 05000334/FIN-2011003-02 | Failure to timely correct Radiation Monitor deficiencies | 2011Q2 | A Green, NRC-identified finding (FIN) was identified in that plans and actions to correct long-standing radiation monitor system instrument deficiencies where not accomplished in a timely manner, in accordance with FENOC CAP procedure NOP-LP2001. Specifically, FENOC failed to correct and return to service radiation monitor instruments for Unit 1 and 2 RSS HX (RM-1RW-100A,B,C,D and 2SWS-RQ100A, B,C,D), in a timely manner, requiring maintenance of alternate monitoring and challenges to assessing radiation detection and assessment during accident situations. This issue was entered into the licensee\'s corrective action program under CR(s) 1191673 and 11-89700. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The inspectors determined that the finding was not similar to the examples for minor deficiencies contained in IMC 0612, Appendix E, Examples of Minor Issues. The finding is more than minor because it affects the Public Radiation Safety cornerstone. The finding is associated with the attribute of plant equipment and instrumentation (process radiation monitors) attribute of the Public Radiation Safety cornerstone to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. In accordance with IMC 0609.04 (Table 3a), Phase 1 - Initial Screening and Characterization of Findings, the finding was evaluated using IMC 0609 Appendix 0, Public Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green) because the finding was not a failure to implement the effluent program or cause any public dose to be exceeded. The cause of this NCV relates to the cross-cutting aspect of Problem, Identification, and resolution, Corrective Action Program, in that FENOC personnel did not take timely corrective actions to develop and implement actions for long-standing radiation monitor deficiencies. |
| 05000334/FIN-2011003-03 | Defective Fuel Injection Pump Supply Lines Provided by the Diesel Engine Manufacturer Results in an Emergency Diesel Generator Being Inoperable | 2011Q2 | On March 25, while Beaver Valley, Unit 2 was shut down in Mode 6 (refueling), scheduled maintenance was performed on the Train A EDG which included replacing the twelve fuel injection pump supply lines with new fuel injection pump supply lines procured from the EDG vendor. During post maintenance testing of the the Train A EDG, the supply line to the number four injector was replaced with its original supply line due to a small fuel leak (approximately 6 drops per minute). Several additional fuel supply line fittings were tightened to eliminate signs of minor leakage and the required post maintenance testing and endurance run of the Train A EDG was completed satisfactorily. After successful post maintenance testing, the Train A EDG was declared operable by the licensee and the Train B EDG was subsequently declared inoperable and removed from service for routine maintenance. Approximately 11 hours later the vendor informed the licensee of its recommendation to remove all of the new fuel injection pump supply lines previously installed on the Train A EDG due to concerns over the adequacy of the assembly method of the fuel line compression fittings. Based on this vendor recommendation, the licensee immediately declared Train A EDG inoperable. Since both EDG trains wer inoperable for approximately 11 hours, during which time no action was immediately taken to restore one to operability as required by TS 3.8.2.B, the licensee was inadvertently in a condition prohibited by TS. When the licensee determined that both EDG trains were inoperable, they entered TS 3.8.2.B, initaed actions to restore one EDG to operable status, and evaluated the issue for reportability and appropriately issued LER 50-412/2011-001, defective fuel injection pump supply lines provided by the diesel engine manufacturer results in an emergency diesel generator being inoperable, dated may 19, 2011. This LER reported that Beaver Valley, unit 2 had been an a condition that was prohibited by TS 3.8.2.B, which requires immediate action if both trains of EDG are inoperble while in Mode 6 operatins. |
| 05000334/FIN-2011005-01 | Unannounced Emergency Response Organization Activation Drill Failure | 2011Q4 | A Green, self-revealing non-cited violation (NCV) of 10 CFR 50.47(b)(2) to ensure timely augmentation of response capabilities is available was identified. Specifically, FENOC failed to fully staff two primary Emergency Response Organization (ERO) positions during an unannounced activation drill. FENOC entered this issue into their corrective action program as CR 2011-04431. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The inspectors determined that the finding was not similar to the examples for minor deficiencies contained in IMC 0612, Appendix E, Examples of Minor lssues. The finding is more than minor because it affects the Emergency Preparedness cornerstone, The finding is associated with the ERO readiness attribute of the Emergency Preparedness cornerstone to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. In accordance with IMC 0609, Appendix B, Sheet 1, Failure to Comply flowchart, the performance deficiency screens to green because it is considered a degraded planning standard function. The cause of this NCV relates to the cross-cutting aspect of Human Performance, Work Practices, in that FENOC personnel did not effectively communicate expectations regarding drill participation and staff did not respond in the required time for ERO positions they had accepted in the call out system. |
| 05000334/FIN-2011007-01 | Failure to Verify the Design Requirements for the Fuel Oil Transfer Pumps | 2011Q2 | The team identified a finding of very low safety significance involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion lll, Design Control because FENOC did not verify or check the adequacy of the Unit 1 emergency diesel generator (EDG) fuel oil transfer system design. Specifically, FENOC did not ensure adequate net positive suction head (NPSH) for the fuel oil transfer pumps during worst case design conditions, and did not evaluate the effect air voids in the suction piping would have on the pumps. FENOC entered the issue into the corrective action program, and performed testing on the fuel oil transfer system and consulted with the pump vendor to determine if the design of the system was adequate. Following completion of the testing and new calculations, FENOC determined that the pumps were operable but degraded. The team determined that the issue was 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 systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of operability or functionality. The team determined that there was not a crosscutting aspect associated with this finding because it was not indicative of current performance. |
| 05000334/FIN-2011007-02 | Inadequate Calculations for Placing SSST LTC in Manual Mode | 2011Q2 | The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion lll, Design Control, because FENOC did not correctly translate the design basis of the electrical distribution system into procedures to ensure operability of offsite power during bus transfers when operating the system service station transformer (SSST) load tap changers (LTC) in the manual mode, an allowed system configuration. Specifically, the team found that procedure\'s supporting calculation did not evaluate the voltage levels on the 480 volt buses. The team determined that during some design basis events, with the tap changer in manual, voltage on the 480 volt vital bus could degrade to a level that would cause the degraded grid relays to trip, resulting in a spurious trip of offsite power. FENOC entered the issue into the corrective action program, and implemented an Operation\'s night order to ensure the LTC was maintained in automatic. The team determined that the issue was 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 systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of operability or functionality. The team determined that there was not a crosscutting aspect associated with this finding because it was not indicative of current performance. |
| 05000334/FIN-2011007-03 | Degraded Voltage Relay Time Delay | 2011Q2 | The team identified an Unresolved ltem (URl) regarding the adequacy of the BVPS degraded voltage protection scheme. The existing degraded voltage relay time delay of 90 +- 5 seconds does not appear to be consistent with the assumption in the UFSAR accident analysis for safety injection flow. The team found that Technical Specification Table 3.3.5-1, Loss of Power Diesel Generator Start and Bus Separation Instrumentation, ltems 3 and 4, lists the degraded voltage relay time delay setpoint as 90 +- 5 seconds. However, the team noted that NRC letter dated June 2, 1977, sent to holders of operating licenses requiring the installation of degraded voltage relays, to ensure safety-related loads had sufficient voltage to respond to an accident, stated in Position 8.1.c that, The allowable time delay, including margin, shall not exceed the maximum time delay that is assumed in the FSAR accident analysis. The NRC safety evaluation report (SER) dated March 3, 1982, concluded that the proposed maximum time delay of 95 seconds does not exceed this maximum time delay. However, the team found that UFSAR Table 14.3.2-8 shows a time delay of < 17 seconds for safety injection flow with offsite power, and < 27 seconds for a loss-of-offsite power/loss-of coolant accident (LOOP/LOCA). The team was concerned that if offsite source voltage was degraded below the level where it was capable of performing its accident mitigation function, but not so low as to actuate the fast acting loss of voltage relays, the time delay assumptions in the accident analysis would not be satisfied. FENOC entered the apparent inconsistency between the design accepted in the 1982 SER and the criteria stated in the 1977 letter into the corrective action program (CR 1 1-95145) for evaluation and resolution. The item was considered unresolved pending NRC review of FENOC actions to address the inconsistency. |
| 05000334/FIN-2011007-04 | Offsite Power Non-Conservative Post Transient Voltage Calculations | 2011Q2 | The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion lll, Design Control, because FENOC did not perform adequate voltage calculations to verify that vital bus voltage levels would be adequate when offsite power was the bus voltage source. The team determined that nonconservative assumptions and evaluations caused the calculation results to predict higher bus voltage levels than could actually occur. Specifically, the team found that FENOC\'s calculational assumptions related to the initial tap position of the SSSTs following bus transfers, evaluation of the effect of the voltage dips that occur during a fast bus transfer, and assumptions for the post event grid voltage condition following the main generator trip could be worse than assumed in the calculation. FENOC entered the issue into the corrective action program, and revised calculations and evaluated post event grid voltage conditions to verify the adequacy of the offsite power source. The team determined that this issue was 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 systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of operability or functionality. The team determined that there was not a crosscutting aspect associated with this finding because it was not indicative of current performance. |
| 05000334/FIN-2011009-01 | Failure to Implement Effective Corrective Actions to Prevent Recurrence of Socket Weld Failures | 2011Q3 | The inspectors identified a Green, self-revealing, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that FENOC failed to take adequate corrective actions to prevent recurrence of a significant condition adverse to quality. Specifically, FENOC\'s extent of condition review and long-term corrective actions following a residual heat removal socket weld failure, caused by vibration-induced high-cycle fatigue, were inadequate to preclude the recurrence of a similar failure on the auxiliary feedwater system. FENOC entered this issue into their corrective action program as condition report 11-01453 for further review. The inspectors determined that FENOC\'s failure to plan or implement adequate corrective actions to prevent recurrence of socket weld failures on safety-related piping was a performance deficiency. This issue was reasonably within FENOC\'s ability to foresee and correct due to previous opportunities to identify and correct socket weld failures on safetyrelated systems at Beaver Valley. The inspectors determined that this self-revealing finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (Le., core damage). The inspectors evaluated the significance of this finding using IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined that this finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of safety system function, and did not screen as potentially risk-significant due to external initiating events. This finding had a cross-cutting aspect in the area of problem identification and resolution because FENOC did not thoroughly evaluate a significant condition adverse to quality such that the resolutions address the extent-of-condition. Specifically, FENOC failed to perform an adequate extent of condition review following the failure of the 1RH-200 socket weld which resulted in not developing adequate corrective actions to address socket welds on the auxiliary feedwater system. |