View source for JAFP-13-0153, LER 12-002-01 for James A. FitzPatrick Nuclear Plant, Regarding High Pressure Coolant Injection Pressure Control Valve Failure

{{Adams
| number = ML14045A274
| issue date = 02/14/2014
| title = LER 12-002-01 for James A. FitzPatrick Nuclear Plant, Regarding High Pressure Coolant Injection Pressure Control Valve Failure
| author name = Coyle L M
| author affiliation = Entergy Nuclear Northeast, Entergy Nuclear Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000333
| license number = DPR-059
| contact person = 
| case reference number = JAFP-13-0153
| document report number = LER 12-002-01
| document type = Letter, Licensee Event Report (LER)
| page count = 6
}}

=Text=
{{#Wiki_filter:Entergy Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.James A.Fitzpatrick NPP P.O.Box 110 Lycoming, NY 13093 Tel 315-349-6024 Fax 315-349-6480 February 14, 2014 JAFP-1 3-0153 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.20555-0001 Lawrence M.Coyle Site Vice President-JAF 

==Subject:==

==Reference:==

==Dear Sir or Madam:==
LER: 2012-002-01, High Pressure Coolant Injection System Pressure Control Valve Failure James A.FitzPatrick Nuclear Power Plant Docket No.50-333 License No.DPR-59 1.Entergy letter to NRC, LER: 2012-002, High Pressure Coolant Injection System Pressure Control Valve Failure, JAFP-12-0132 dated October 29, 2012 This letter submits a revision to LER: 2012-002 submitted by letter dated October 29, 2012[Reference 1].Reference 1 was submitted to report 10 CFR 50.73(a)(2)(v)(D),"Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident."
This revision adds criterion 10 CFR 50.73(a)(2)(i)(B),"Any operation or condition which was prohibited by the plant's Technical Specifications."
There are no commitments contained in this report.Questions concerning this report may be addressed to Mr.Chris Adner, Manager, Regulatory Assurance at (31~, 349-6766.LMC/CMA/ds Enclosure(s):
LER: 2012-002-01, High Pressure Coolant Injection System Pressure Control Valve Failure cc: USNRC, Region 1 USNRC, Project Directorate USNRC Resident Inspector INPO Records Center Sincerely,A.
..nce M.Coyle Site Vice President NRC FORM 366 (01-2014)NRC FORM 366U.S. NUCLEAR REGULATORY COMMISSION(01-2014)LICENSEE EVENT REPORT (LER)(See Page 2 for required number ofdigits/characters for each block)APPROVED BY OMB: NO. 3150-0104                             EXPIRES: 01/31/2017Estimated burden per response to comply with this mandatory collection request: 80 hours.Reported lessons learned are incorporated into the licensing process and fed back to industry.Send comments regarding burden estimate to the FOIA, Privacy and Information CollectionsBranch (T-5 F53),  U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,  or byinternet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information andRegulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC20503. If a means used to impose an information collection does not display a currently valid OMBcontrol number, the NRC may not conduct or sponsor, and a person is not required to respond to,the information collection.1. FACILITY NAMEJames A. FitzPatrick Nuclear Power Plant2. DOCKET NUMBER050003333. PAGE1  OF  54. TITLE High Pressure Coolant Injection Pressure Control Valve Failure 5.EVENTDATE 6.LERNUMBER 7.REPORTDATE 8.OTHERFACILITIES INVOLVEDMONTH DAYYEARYEARSEQUENTIALNUMBERREVNO.MONTHDAYYEARFACILITY NAME N/A   DOCKET NUMBER N/A08 30 20122012-002-0102 14 2014FACILITY NAME N/A   DOCKET NUMBER N/A 9.OPERATING MODE 11.THIS REPORT ISSUBMITTEDPURSUANT TOTHEREQUIREMENTS OF 10 CFR&sect;:(Checkallthatapply)1  20.2201(b)  20.2203(a)(3)(i)  50.73(a)(2)(i)(C)  50.73(a)(2)(vii)  20.2201(d)  20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A)  50.73(a)(2)(viii)(A)  20.2203(a)(1)  20.2203(a)(4)  50.73(a)(2)(ii)(B)  50.73(a)(2)(viii)(B)  20.2203(a)(2)(i)  50.36(c)(1)(i)(A)  50.73(a)(2)(iii)  50.73(a)(2)(ix)(A)10. POWER LEVEL  20.2203(a)(2)(ii)  50.36(c)(1)(ii)(A)  50.73(a)(2)(iv)(A)  50.73(a)(2)(x)100  20.2203(a)(2)(iii)  50.36(c)(2)  50.73(a)(2)(v)(A)  73.71(a)(4)  20.2203(a)(2)(iv)  50.46(a)(3)(ii)  50.73(a)(2)(v)(B)  73.71(a)(5)  20.2203(a)(2)(v)  50.73(a)(2)(i)(A)  50.73(a)(2)(v)(C)  OTHER  20.2203(a)(2)(vi)  50.73(a)(2)(i)(B)  50.73(a)(2)(v)(D)Specify in Abstract below or inNRC Form 366A 12.LICENSEE CONTACTFORTHIS LERFACILITY NAMEMr. Chris M. Adner, Regulatory Assurance ManagerTELEPHONE NUMBER(Include Area Code) 3153496766 13.COMPLETEONELINE FOREACHCOMPONENTFAILUREDESCRIBED INTHISREPORTCAUSESYSTEMCOMPONENTMANU-FACTURERREPORTABLETO EPIXCAUSESYSTEMCOMPONENTMANU-FACTURERREPORTABLETO EPIX D BJ PCVM120 N14. SUPPLEMENTAL REPORT EXPECTED   YES(If yes, complete 15. EXPECTED SUBMISSION DATE)   NO15. EXPECTEDSUBMISSIONDATEMONTHDAYYEARABSTRACT(Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)On August 28, 2012, while testing the High Pressure Coolant Injection System (HPCI), it was discovered that water was leaking intothe reactor building sump (20TK-69A). This leakage was due to the lifting of HPCI booster pump recirculation safety valve (23SV-66)caused by a failure of HPCI booster pump P-1B recirculation pressure control valve (23PCV-50).Due to the amount of leakage, the HPCI system may not have been able to meet its mission time without realigning its suction sourceto the torus. As a result, HPCI was declared inoperable. The most probable cause of the 23PCV-50 failure was material introducedinto the sensing line and sensing line filter when the system was filled using torus water during a maintenance activity on June 8, 2012.Immediate corrective actions included replacing the snubber and filter, refilling the sensing line with condensate storage tank water,and then venting the system. Future corrective actions will include revising the procedure used to fill the HPCI system such that itcontains additional guidance when filling the HPCI suction piping.This event is reportable in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented the fulfillmentof a safety function of a system required to mitigate the consequences of an accident, and 10 CFR 50.73(a)(2)(i)(B), any operation orcondition which was prohibited by the plant's Technical Specifications.
NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION(01-2014)LICENSEE EVENT REPORT (LER)     CONTINUATION SHEETAPPROVED BY OMB: NO. 3150-0104                             EXPIRES: 01/31/2017Estimated burden per response to comply with this mandatory collection request: 80 hours.Reported lessons learned are incorporated into the licensing process and fed back to industry.Send comments regarding burden estimate to the FOIA, Privacy and Information CollectionsBranch (T-5 F53),  U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,  or byinternet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information andRegulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC20503. If a means used to impose an information collection does not display a currently valid OMBcontrol number, the NRC may not conduct or sponsor, and a person is not required to respond to,the information collection.1. FACILITY NAMEJames A. FitzPatrick Nuclear Power Plant2. DOCKET 050003336. LER NUMBER3. PAGE2  OF  5YEARSEQUENTIAL NUMBERREV NO.2012 002 01NRC FORM 366A (01-2014)NARRATIVEBACKGROUNDOn January 18, 1988, a design change was made that installed a larger inline filter in the pressure sensing linefor the High Pressure Coolant Injection (HPCI) [EIIS System Identifier:  BJ] booster pump P-1B recirculationpressure control valve (23PCV-50) [EIIS Component Identifier:  PCV]. This change was made because ofseveral instances where the 23PCV-50 filter or snubber would become blocked by debris thereby preventing thepressure control valve from controlling. A two year preventative maintenance (PM) activity was also establishedto clean, inspect, and replace the filter and snubber.On April 30, 2012, a new revision of OP-15, "High Pressure Coolant Injection" was issued. This revision of OP-15 added a new section, G.9, "Fill and Vent HPCI Suction Piping From Condensate Storage Tanks (CST)," toaddress a corrective action identified during the Nuclear Regulatory Commission (NRC) inspection on gasaccumulation earlier in the year.On June 8, 2012, a HPCI outage was conducted in order to perform PM on HPCI Booster Pump P-1B SuctionFrom Suppression Pool Check Valve (23HPCI-61) [EIIS Component Identifier:  V]. This required the HPCIsystem to be isolated and drained, including the pump and suction line piping. In addition, the filter and snubberon the pressure sensing line for 23PCV-50 were also replaced as required by the PM.During restoration, a portion of the HPCI suction piping was filled and vented from the torus per OP-15, SectionG.8, "Fill and Vent HPCI Suction Piping from Torus."  The remaining HPCI suction piping was filled and ventedfrom the CSTs in accordance with OP-15, Section G.9. Post work and return to service testing was completedsatisfactory three days later and operability was demonstrated by a successful completion of ST-4N, "HPCIQuick-Start, Inservice, and Transient Monitoring Test (IST)."EVENT DESCRIPTION & ANALYSISOn August 28, 2012, while running the HPCI turbine for ST-4N, several annunciators were received in thecontrol room, indicating that the reactor building equipment sump "A" (20TK-69A) [EIIS Component Identifier:TK] was being overflowed and water was running down into the floor sump. This condition was confirmedvisually by an operator. At that time the source of the extra water was unknown. Since the volume of waterentering 20TK-69A was greater than what was expected to come from the HPCI system. It was assumed thattorus water was coming through a leaking check valve on the discharge of the reactor building equipment drainsump pump. At the time of discovery, torus water level was being lowered by pumping it to the radwaste system[EIIS System Identifier:  WD] via the equipment drain discharge header.On August 30, 2012, operators performed ST-4E, "HPCI and SGT Logic System Functional and SimulatedAutomatic Actuation Test."  The data collected during this surveillance revealed that while the HPCI turbine wasin operation, there was approximately 75 gpm of water going into the "A" reactor building sump. The source ofthis water was determined to come from HPCI Booster Pump P-1B Recirculation Safety Valve (23SV-66) [EIISComponent Identifier:  RV] which was lifting on high pressure. Troubleshooting determined that the cause of23SV-66 to lift was a failure of 23PCV-50 to properly control pressure.
NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION(01-2014)LICENSEE EVENT REPORT (LER)CONTINUATION SHEET1. FACILITY NAME2. DOCKET6. LER NUMBER3. PAGEJames A. FitzPatrick Nuclear Power Plant05000333YEARSEQUENTIALNUMBER REVN0.3 OF 52012-002- 01NRC FORM 366A (01-2014)Control pressure for 23PCV-50 is 75 psia which is the design pressure for the HPCI lube oil cooler (23E-2) [EIISComponent Identifier:  CLR] and gland seal condenser (23E-1) [COND]. However, data collected during the ST-4E run on August 30, 2012, demonstrated that 23PCV-50 was not repositioning as expected. The increaseddown stream pressure caused 23SV-66 to lift, allowing CST water into the reactor building equipment sump.The HPCI system is considered operable when it is aligned to one or both CSTs with power available to supportautomatic realignment to the suppression pool if required. This is based on the design of the CSTs and theaccident analysis which credits the suppression pool for supplying the HPCI System. With an assumed leakageof 75 gpm of CST water being directed into 20TK-69A, HPCI may not have been able to meet its mission timewithout realigning its suction to the torus. If the HPCI suction source re-aligned to the torus, due to low CSTlevels, the water being discharged to 20TK-69A would be torus water; this condition would result in total leakagesources outside containment exceeding the 5 gpm limit established by the Final Safety Analysis Report (FSAR).As a result, HPCI was declared inoperable on August 30, 2012. On September 2, 2012, after replacing thesensing line filter and snubber; flushing the system with clean CST water; and successfully performing return toservice testing; HPCI was restored to operable status. This was reported to the NRC on August 30, 2012, viaENS #48258. It is being reported in this LER in accordance with 10 CFR 50.73(a)(2)(v)(D), any event orcondition that could have prevented the fulfillment of the safety function of structures or systems that are neededto mitigate the consequences of an accident.The action that led to 23PCV-50 failing to control pressure was filling the HPCI pump suction from the torusduring restoration from the June 2012 maintenance outage. That action introduced foreign material into thesuction line that eventually fouled the filter in the 23PCV-50 sensing line. The date that 23PCV-50 was notcapable of performing its function is indeterminate. The period of time from June 2012 to August 2012 is greaterthan the allowed outage time for HPCI (TS LCO 3.5.1), therefore this event is also being reported under 10 CFR50.73(a)(2)(i)(B), any operation or condition which was prohibited by the plant's Technical Specifications.CAUSE OF EVENTMechanisticThe apparent cause of the event was determined to be material in the 23PCV-50 sensing line and filter. Thiswas validated by physical inspection during troubleshooting. The material was a result of filling and venting the23PCV-50 sensing line with torus water containing suspended solids.Normally the 23PCV-50 sensing line is maintained full of water. With its short stroke, suspended solids don'tmake their way up the line and into the filter. However, during the HPCI LCO in June, both the HPCI systemand the 23PCV-50 pressure sensing lines were drained at the same time. Therefore, when the HPCI suctionpiping was filled from the torus, the sensing line was also filled. This resulted in suspended solids from the toruswater clogging the filter in the sensing line.ProgrammaticThe event was reviewed for organizational and programmatic deficiencies that may have caused or contributedto the event. It was determined that Operations Procedure, OP-15 had insufficient detail in its guidance for fillingand venting from the torus. This had the unintended consequence of filling portions of the HPCI line, includingthe instrument line for 23PCV-50, with material from the torus.
NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION(01-2014)LICENSEE EVENT REPORT (LER)CONTINUATION SHEET1. FACILITY NAME2. DOCKET6. LER NUMBER3. PAGEJames A. FitzPatrick Nuclear Power Plant05000333YEARSEQUENTIALNUMBER REVN0.4 OF 52012-002- 01NRC FORM 366A (01-2014)EXTENT OF CONDITIONAn extent of condition review was performed for other PCVs subject to the same failure mode. The systemsreviewed were HPCI, Reactor Core Isolation Cooling [EIIS System Identifier:  BN], Residual Heat Removal [EIISSystem Identifier:  BO], and Core Spray [EIIS System Identifier:  BM]. This review did not identify any otherPCV that was applicable to the failure mode described in this LER.FAILED COMPONENT IDENTIFICATIONDescription:  HPCI Booster Pump P-1B Recirc Pressure Control ValveManufacturer:  Masoneilan Intl, Inc.Model/Part Number:  525NPRDS Manufacturer Code:  M120FitzPatrick Component ID:  23PCV-50CORRECTIVE ACTIONSCompleted The PCV in-line filter and snubber have been replaced. The pressure sensing line for 23PCV-50 was flushed with clear water. 23PCV-50 was tested satisfactory. All other components in the HPCI system have been evaluated for extent of condition, and are notsusceptible to this failure mode. HPCI system has been tested successfully per ST-4N.Future Actions Revise OP-15 to add additional guidance for filling the HPCI suction piping. Evaluate a design change to have 23SV-66 discharge into torus vice equipment sump. Revise PM to fill sensing line using a clean water source.ASSESSMENT OF SAFETY CONSEQUENCESThe HPCI System is designed to provide adequate core cooling to limit fuel clad temperatures in the event of asmall break in the Reactor Coolant System piping with a loss of coolant that does not result in rapiddepressurization of the reactor pressure vessel (RPV).The significance of this condition is based on the safety function performed by the HPCI system. With 23PCV-50 not controlling pressure, 23SV-66 would lift continuously with HPCI in operation. This would initially result inCST water being directed to 20TK-69A. If the HPCI suction source re-aligned to the torus, due to low CSTlevels, the water being discharged to 20TK-69A would be torus water; this condition would result in total leakagesources outside containment exceeding the 5 gpm limit established by the FSAR. An assessment of thepotential risk contributions associated with this alignment was performed, and the result screened, per thecriteria established in NRC IMC 0609, as very low safety significance with nominal risk.Radiological & Industrial SafetyThere were no actual radiological or industrial safety consequences. The potential impact on radiological andindustrial safety is minimal; Secondary Containment is required to be Operable in all applicable Modes ofoperation.
NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION(01-2014)LICENSEE EVENT REPORT (LER)CONTINUATION SHEET1. FACILITY NAME2. DOCKET6. LER NUMBER3. PAGEJames A. FitzPatrick Nuclear Power Plant05000333YEARSEQUENTIALNUMBER REVN0.5 OF 52012-002- 01NRC FORM 366A (01-2014)Nuclear SafetyThere was no actual or potential nuclear safety consequences associated with this condition. At all times HPCIwas available to provide a source of RPV water inventory in the event of a loss of coolant accident. Additionally,the Automatic Depressurization System (ADS) in combination with the Low Pressure Coolant Injection system(LPCI) and the Core Spray system (CS) were available to provide core cooling during the period of HPCIinoperability. Secondary containment also remained OPERABLE during this period.This deficiency did have a potential impact on the Primary Coolant Sources Outside Containment Programrequired by TS 5.5.2. This program is in place to ensure that leaks are tracked, assessed, and prioritized suchthat the potential to exceed post accident release rates are minimized. With respect to this program, it wouldonly be impacted in the event that the HPCI suction was aligned to the torus.The potential impact of this condition was minimized because during the course of this event, the HPCI systemsuction was aligned to the CST's. In addition, Emergency Operating Procedures (EOP) preferentially maintainthe HPCI suction aligned to the CSTs and during accident conditions the EOPs do address high water levels inthe reactor building sumps and crescents; actions include isolating systems discharging into the area, shuttingdown the reactor and depressurizing the reactor.SIMILAR EVENTSInternal operating experience (OE) was reviewed through Entergy's corrective action program. There were norelevant events found. Similarly, external industry OE was reviewed via INPO. Although there were severalevents that had some applicability to JAF, none of the events were relevant with regards to the event beingreported in this LER. Insights from the OE search were incorporated into the corrective action plan.REFERENCES JAF Condition Reports:  CR-JAF-2012-04994, CR-JAF-2012-04958, CR-JAF-2012-03015 JAF TS 3.5.1, ECCS - Operating, TS 5.5.2 - Primary Coolant Sources Outside Containment JAF Engineering Change 39479 JAF FSAR 6.4.1 High Pressure Coolant Injection System}}