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{{Adams
#REDIRECT [[05000293/LER-2013-005]]
| number = ML13211A174
| issue date = 07/22/2013
| title = LER 13-005-00 and LER 13-006-00 for Pilgrim Regarding Primary Containment Declared Inoperable During HPCI Testing & HPCI Controller Failure to Achieve Rated Flow While in Auto Mode
| author name = Noyes D
| author affiliation = Entergy Nuclear Operations, Inc
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000293
| license number = DPR-035
| contact person =
| case reference number = 2.13.058
| document report number = LER 13-005-00, LER 13-006-00
| document type = Letter, Licensee Event Report (LER)
| page count = 12
}}
 
=Text=
{{#Wiki_filter:SEntergyEntergy Nuclear Operations, Inc.600 Rocky Hill RoadPlymouth, MA 02360Pilgrim Nuclear Power StationJuly 22, 2013U.S. Nuclear Regulatory Commission Attn: Document Control DeskWashington, D.C. 20555
 
==SUBJECT:==
Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power StationDocket No.: 50-293License No.: DPR-35Licensee Event Report 2013-005-00, Primary Containment Declared Inoperable During HPCI TestingLicensee Event Report 2013-006-00, HPCI Controller Failure to Achieve Rated Flowwhile in Auto ModeLETTER NUMBER: 2.13.058
 
==Dear Sir or Madam:==
The enclosed Licensee Event Reports are submitted in accordance with 10 CFR 50.73.LER 2013-005-00, "Primary Containment Declared Inoperable During HPCI Testing"LER 2013-006-00, "HPCI Controller Failure to Achieve Rated Flow while in Auto Mode"This letter contains no commitments.
Please do not hesitate to contact Mr. Joseph R. Lynch, (508) 830-8403, if there are any questions regarding this submittal.
Sincerely, David No esDirector, Nuclear Safety Assurance DN/WGLAttachment 1: Licensee Event Report 2013-005-00, Primary Containment Declared Inoperable DuringHPCI Testing (4 pages)Attachment 2: Licensee Event Report 2013-006-00, HPCI Controller Failure to Achieve Rated Flow whilein Auto Mode (4 Pages)
PNPS Letter 2.13.058Page 2 of 2cc:Mr. William M. DeanRegional Administrator, Region 1U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd., Suite 100King of Prussia, PA 19406-1415 INPO Records700 Galleria ParkwayAtlanta, GA 30399-5957 Mr. Richard V. Guzman, Project ManagerDivision of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8-C211555 Rockville PikeRockville, MD. 20852USNRC Senior Resident Inspector Pilgrim Nuclear Power Station Attachment ILetter Number 2.13.058Licensee Event Report 2013-005-00 Primary Containment Declared Inoperable During HPCI Testing(4 Pages)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2013 (10-2010)
Estimated 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 FOINPrivacy Service Branch (T-5 F53), U.S.Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail toLICENSEE EVENT REPORT (LER) infocollects.resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory
: Affairs, NEOB-10202, (3150-0104),
Office of Management and Budget, Washington, DC 20503.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 NAME 2. DOCKET NUMBER 3. PAGEPilgrim Nuclear Power Station 05000293 1 OF 44. TITLEPrimary Containment Declared Inoperable During HPCI Testing5. EVENT DATE 6. LER NUMBER 7. REPORT DATE 8. OTHER FACILITIES INVOLVEDRE FACILITY NAME DOCKET NUMBERMONTH DAY YEAR YEAR SEQUENTIAL V MONTH DAY YEAR N/ANUMBER N0FACILITY NAME DOCKET NUMBER05 23 2013 2013 005 00 7 22 2013 NIA9. OPERATING MODE 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)[ 20.2201(b) 20.2203(a)(3)(i)
[ 50.73(a)(2)(i)(C)
[] 50.73(a)(2)(vii)
N L 20.2201(d) 20.2203(a)(3)(ii) r 50.73(a)(2)(ii)(A)
E] 50.73(a)(2)(viii)(A)
H 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B)
[] 50.73(a)(2)(viii)(B)
E 20.2203(a)(2)(i)
[ 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A) 20.2203(a)(2)(ii)
[ 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x)
: 10. POWER LEVEL L- 20.2203(a)(2)(iii)
-50.36(c)(2)
[ 50.73(a)(2)(v)(A) 73.71 (a)(4)E 20.2203(a)(2)(iv)
-50.46(a)(3)(ii)
LI 50.73(a)(2)(v)(B) 73.71 (a)(5)20.2203(a)(2)(v) 50.73(a)(2)(i)(A)
F] 50.73(a)(2)(v)(C)
E] OTHER2% 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B)
F] 50.73(a)(2)(v)(D)
Specify in Abstract below or inNRC Form 366A12. LICENSEE CONTACT FOR THIS LERNAME TELEPHONE NUMBER (Include Area Code)Joseph R. Lynch, Licensing Manager 1(508)-830-8403
: 13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORTCAUSE SYSTEM COMPONENT MANU REPORTABLE CAUSE SYSTEM COMPONENT MANU- REPORTABLE FACTURER TO EPIX FACTURER TO EPIXA BJ N/A N/A Y A NH N/A N/A Y14. SUPPLEMENTAL REPORT EXPECTED
: 15. EXPECTED MONTH DAY YEARSUBMISSION Yes (If yes, complete
: 15. EXPECTED SUBMISSION DATE NNO DATEABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)At 0455 hours on Thursday, May 23, 2013, with Pilgrim Station in the Startup/Hot Standby Mode and reactorpressure approximately 550 psig, primary containment was declared inoperable due to a leak in the HighPressure Coolant Injection System (HPCI) turbine exhaust line observed while performing the HPCI systemflow test. Power ascension was suspended pending investigation and repair. Pilgrim entered into Technical Specification 3.7.A.2 requiring the plant to be in cold shutdown within 24 hours. All other safety systemsfunctioned as required.
The apparent cause of the leak in the HPCI turbine exhaust line was failure to adequately tighten all of theflange bolting due to unique bolting and flange configuration associated with the new check valve andbutterfly valve installation during the refueling outage (RFO)-19.
The flange bolting was subsequently re-tightened applying the target torque values for the application.
A Type B Local Leak Rate Test wasperformed on the butterfly valve outlet flange and the measured leak rate met the Technical Specification and10 CFR 50, Appendix J requirements, and primary containment was declared operable.
This event posed no threat to public health and safety.NRC FORM 366 (10-2010)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)CONTINUATION SHEET1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGESEQUENTIAL REVPilgrim Nuclear Power Station 05000293 YA NUMBER NO. 1 ,2 OF 42013- 005 -00NARRATIVE BACKGROUND:
During RFO-19, stop check valve 2301-74 in the HPCI turbine exhaust line was replaced by a swing check valveand a new butterfly valve (23-HO-321) was installed at the outlet of 2301-74.
The removed valve and the newvalves are all flanged-end valves. As part of the plant startup following RFO-1 9, the HPCI system was tested at150 psig reactor pressure in accordance with PNPS procedure 8.5.4.3, "High Pressure Coolant Injection Operability Demonstration and Flow Rate Test at 150 psig"; conduct of this procedure was also one of the postwork test requirements for the valve replacement modification.
During HPCI operation, water leakage wasobserved from the outlet flange on valve 23-HO-321.
The HPCI turbine exhaust steam is discharged to the torus through two check valves, 2301-45 and 2301-74,both of which are primary containment isolation valves. The HPCI turbine exhaust check valve 2301-45 islocated closest to the turbine, and valve 2301-74 is located downstream closest to the torus. These valves aretested locally for leak tightness in accordance with the station's local leak rate testing (LLRT) procedure, PNPSprocedure 8.7.1.5, "Local Leal Rate testing of Primary Containment Penetrations, Isolation Valves, andInspection of Containment",
to meet the requirements of 10CFR50 Appendix J. The purpose of the Appendix Jis to ensure the integrity of the primary containment to contain any releases of radioactive material tocontainment inside the primary containment.
The piping system flanges are also tested as part of the LLRTprogram.Technical Specification 3.7.A. 2 requires primary containment integrity at all times when the reactor is critical orwhen the reactor coolant temperature is above 212'F and fuel is in the reactor vessel. To assure primarycontainment integrity, all containment isolation valves must be operable or closed and pressure boundary mustremain intact to comply with radiological release limits specified in 10 CFR 100 in the event of a break in theprimary coolant system piping.EVENT DESCRIPTION:
On May 23, 2013, the HPCI system flow test was performed in accordance with PNPS 8.5.4.3 with reactorpressure at 150 psig and Pilgrim Station in the Startup/Hot Standby Mode. During the test, water was observedleaking from the butterfly valve outlet-to-plant piping flange on the HPCI turbine exhaust piping. The water wasfrom condensed steam in the line. Because the leakage indicated there was a leak path past both o-ring sealsat this flange joint, Operations made the conservative decision to declare primary containment inoperable.
Pilgrim entered into the cold shutdown LCO on May 23, 2013, at 0455 and upon completing the repair, exitedthe LCO on May 23, 2013, at 1822.CAUSE OF THE EVENT:The cause was the failure to provide work package instructions necessary to adequately tighten all of the boltingassociated with the affected HPCI turbine exhaust piping flange joint. This was due to the lack of understanding of the joint bolting configuration that 4 of the studs were threaded into each side of the butterfly valve body,which is different from the 16 studs that pass through the butterfly valve flange bolt holes and are captured bynuts at the adjacent check valve and plant piping flanges.There were no component failures.
NRC FORM 366A (10-2010)
NRC FORM 366A(10-2010)
U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)CONTINUATION SHEET1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGESEQUENTIAL REVPilgrim Nuclear Power Station 05000293 Y NUMBER NO. 3 OF 42013- 005 -00CORRECTIVE ACTIONS:The following corrective actions were implemented:
Once appropriate instructions were provided, the flange bolting was tightened to the target torque value of 700ft-lb.The HPCI system was subsequently operated in accordance with PNPS 8.5.4.3 with reactor pressure atapproximately 550 psig. A Type B LLRT test was performed on the butterfly valve outlet flange and themeasured leak rate met the requirements of PNPS procedure 8.7.1.5, 10CFR50 Appendix J, and the Technical Specifications.
The primary containment was then declared operable.
Additional corrections are captured in the corrective action program under Condition Report, CR-PNP-2013-04262.ASSESSMENT OF SAFETY CONSEQUENCES:
This condition posed no threat to the public health and safety.The event occurred during power ascension from RFO-19. Core Thermal Power was at approximately 2% andreactor pressure was approximately 550 psig.The safety objective of Primary Containment is to limit the release of fission products in the event of a designbasis accident so that off-site doses would not exceed the requirements of 10 CFR Part 100 and to preventexcessive fuel cladding temperatures.
Primary containment consists of the drywell and the pressure suppression chamber (torus).
The torus provides the water supply for the Core Standby Cooling Systems.The HPCI exhaust piping leak identified in this report is located in the HPCI Room. The HPCI Room is locatedoutside of primary containment but inside the Reactor Building or Secondary Containment.
Secondary Containment is provided to minimize ground level release resulting from potential leaks from primarycontainment.
The HPCI exhaust line connects to the Torus and is submerged below water level in the Torus. A vacuumbreaker line is provided to prevent condensing steam in the exhaust line from drawing a vacuum and drawingwater into to exhaust line which would result in water hammer. There are two motor operated valves in thevacuum breaker line.During the time period that the primary containment barrier was degraded, Secondary Containment wasOperable as was the motor operated valves in the vacuum breaker, which were able to isolate the exhaust lineleak. In addition, the ADS, CS, RHR, and RCIC systems were either operable or available.
These systems andcomponents provided capability to isolate the identified leak path and supply makeup water to the vessel toensure adequate core cooling.The leak was repaired and there was no long term safety significance associated with the event.REPORTABILITY:
NRC FORM 366A (10-2010)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)CONTINUATION SHEET1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGEYAT SEQUENTIAL REVPilgrim Nuclear Power Station 05000293 NUMBER NO. 4 OF 42013- 005 -00This occurrence was reported to the USNRC in accordance with 10 CFR 50.72(b)(3)(ii)(A) and 10 CFR50.72(b)(3)(v)
(C) and (D) as documented in EN# 49061.PREVIOUS OCCURRENCES:
There were no previous events related to the HPCI turbine exhaust piping flange leakage.ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES:The EIIS codes for Components and Systems referenced in this report are as follows:COMPONENTS CODESPipe Fittings PSFValve, Isolation ISVSYSTEMSHigh Pressure Coolant Injection (HPCI) BJReactor Containment Building (BWR Primary NHContainment
 
==REFERENCES:==
 
Condition Report, CR-PNP-2013-04262, HPCI Turbine Exhaust Piping Flange LeakNRC FORM 366A (10-2010)
Attachment 2Letter Number 2.13.058Licensee Event Report 2013-006-00 HPCI Controller Failure to Achieve Rated Flow while in Auto Mode(4 Pages)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES:
10/31/2013 (10-2010)
Estimated 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 FOlNPrivacy Service Branch (T-5 F53), U.S.Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail toLICENSEE EVENT REPORT (LER) infocollects.resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory
: Affairs, NEOB-10202, (3150-0104),
Office of Management and Budget, Washington, DC 20503.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 NAME 2. DOCKET NUMBER 3. PAGEPilgrim Nuclear Power Station 05000293 1 OF 44. TITLEHPCI Flow Controller Failure to Achieve Rated Flow while in Auto Mode5. EVENT DATE 6. LER NUMBER 7. REPORT DATE 8. OTHER FACILITIES INVOLVEDRE FACILITY NAME DOCKET NUMBERMONTH DAY YEAR YEAR SEQUENTIAL V MONTH DAY YEAR 05000NUMBER NFACILITY NAME DOCKET NUMBER05 23 2013 2013 006 00 7 22 2013 050009. OPERATING MODE 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)L 20.2201(b) 20.2203(a)(3)(i)
[] 50.73(a)(2)(i)(C) 50.73(a)(2)(vii)
N 20.2201 (d) 20.2203(a)(3)(ii)
E] 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A)
H 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)20.2203(a)(2)(ii)
] 50.36(c)(1)(ii)(A)
H 50.73(a)(2)(iv)(A) 50.73(a)(2)(x)
: 10. POWER LEVEL 20.2203(a)(2)(iii)
H 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)
H 73.71 (a)(5)0 % 20.2203(a)(2)(v)
H 50.73(a)(2)(i)(A)
H 50.73(a)(2)(v)(C)
H OTHER002% 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)
Specify in Abstract below or inNRC Form 366A12. LICENSEE CONTACT FOR THIS LERNAME TELEPHONE NUMBER (Include Area Code)Joseph R. Lynch, Licensing Manager 1 (508)-830-8403
: 13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORTCAUSE SYSTEM COMPONENT MANU- REPORTABLE CAUSE SYSTEM COMPONENT MANU- REPORTABLE FACTURER TO EPIX FACTURER TO EPIXX BJ FIC N430 Yes14. SUPPLEMENTAL REPORT EXPECTED
: 15. EXPECTED MONTH DAY YEARSUBMISSION Yes (If yes, complete
: 15. EXPECTED SUBMISSION DATE NO DATEABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)On May 23, 2013, at 1050 hours, during plant start up from Refueling Outage (RFO-19) with the reactor at2% core thermal power, reactor pressure at -525 psig, and the mode switch in the Startup / Hot Standbyposition, Pilgrim Nuclear Power Station (PNPS) declared the High Pressure Coolant Injection (HPCI) systeminoperable due to failure of the HPCI flow indicating controller (FIC-2340-1) to maintain system discharge flow rate above 4250 gpm while in the automatic mode from the Main Control Room during planned postmaintenance testing.
Limiting Condition for Operation (LCO) actions for Technical Specification (TS) 3.5.C.2were entered.The cause of the event was determined to be FIC-2340-1 out of calibration due to degradation of theautomatic (null) control/output circuit.
Corrective action was taken to troubleshoot, recalibrate and adjust theflow controller.
Post work testing verified HPCI flow controller operability.
This event was not risk significant and had no adverse impact on the health and/ or safety of the public.NRC FORM 366 (10-2010)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)CONTINUATION SHEET1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGEYEAR SEQUENTIAL REVPilgrim Nuclear Power Station 05000293 NUMBER NO. 2 OF 42013 -006 -00NARRATIVE BACKGROUND:
The Pilgrim Station Core Standby Cooling systems (CSCS) consist of the High Pressure Coolant Injection (HPCI) system, Automatic Depressurization system (ADS), Residual Heat Removal (RHR) System LowPressure Injection (LPCI) mode, and Core Spray (CS) system. The HPCI system is designed to pump waterinto the reactor vessel for high pressure core cooling.
Although not part of the CSCS, the reactor core isolation cooling (RCIC) system is also designed to pump water into the reactor vessel for high pressure core cooling,The HPCI System flow indicating controller (FIC-2340-1) installed in the Main Control Room functions tomaintain a process flow at a desired set-point.
The controller provides for both manual and automatic processcontrol and has an internal set-point control circuit.
The controller compares a process variable (HPCI flow fromFT-2358) with a control set-point (normally set at 4250 gpm).Engineering Change EC12967 was issued to replace obsolete GMAC HPCI flow controllers with NUSInstrument Corporation Model PID901-540 flow controllers.
These NUS flow controllers were reverseengineered and intended to be equivalent replacements.
On 2/24/13 the NUS HPCI System flow controller wasinstalled in the Main Control Room and successfully tested to verify operability.
On May 23, 2013, the plant was starting up from a Refueling Outage (RFO-19).
In accordance with Technical Specification (TS) 3.5.C.1, the HPCI System is required to be tested at a reactor pressure of 150 psig to verifysystem operability.
Procedure 8.5.4.3 provides test criteria for system operability and ensures that the systemautomatically starts and can control flow at or above 4250 gpm. The HPCI system was operated on May 23,2013 at 0034 hours and met test criteria.
Subsequent HPCI system runs were planned to address postmaintenance test requirements.
EVENT DESCRIPTION:
On May 23, 2013, at 1050 hours, during plant start-up from RFO-19 with the reactor at 2% core thermal power,reactor pressure at -525 psig, and the mode switch in the Startup/
Hot Standby position, PNPS declared theHPCI system inoperable due to failure of the HPCI flow indicating controller to maintain system discharge flowrate above 4250 gpm while in the automatic mode from the Main Control Room during planned postmaintenance testing.
Limiting Condition for Operation (LCO) actions for Technical Specification 3.5.C.2 wereentered.CAUSE OF THE EVENT:The apparent cause evaluation identified that the direct cause of the HPCI system failure was flow controller FIC-2340-1 out of calibration by 550 gpm due to degradation of the flow controller automatic (null) control/output circuit.The apparent cause evaluation was based on removal of the flow controller, bench testing, and implementing adetailed troubleshooting plan.NRC FORM 366A (10-2010)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)CONTINUATION SHEET1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGEPilgrim Nuclear Power Station YEAR SEQUENTIAL REV05000293
, NUMBER NO. 3 OF 42013 -006 -00EXTENT OF CONDITION:
PID901-540 flow controllers are installed in the HPCI system, RCIC system, and Control Rod Drive (CRD)system. Based on system testing, the condition was only identified in the HPCI flow controller located in theMain Control Room.CORRECTIVE ACTIONS:Corrective actions completed included troubleshooting, bench testing, and successful recalibration andadjustment of the HPCI flow controller.
Post calibration testing confirmed stable operation of the flow controller during HPCI System Operability test runs.Corrective actions planned include:-Replace the installed Main Control Room HPCI flow controller.
-Send the replaced flow controller to the vendor\manufacturer for evaluation.
-Revise flow controller calibration procedures as necessary to address adequate guidance/steps to check forthe degradation that caused this event.-After vendor evaluation, incorporate appropriate revisions into applicable procedures and document actions inthe Corrective Action Program (CAP).These corrective actions will be tracked in the Corrective Action Program via CR-PNP-2012-4286.
ASSESSMENT OF SAFETY CONSEQUENCES:
The event occurred during power ascension from RFO-1 9. Core Thermal Power was at approximately 2% andreactor pressure was approximately 525 psig.CSCS systems include HPCI, ADS, CS, and RHR -LPCI mode. Although not part of the CSCS systems, theRCIC system is capable of providing water to the reactor vessel for high pressure core cooling, similar to theHPCI system.During the time period that HPCI flow controller was out of service, the ADS, CS, RHR, and RCIC systems wereeither operable or available.
These systems provided capability to supply makeup water to the vessel andensured adequate core cooling while the HPCI system was not operable.
During the event, the HPCI systemautomatically started and controlled flow at slightly less than 4250 gpm. HPCI system was restored to operablestatus and there was no long term safety significance associated with the event.The bounding case of risk assessment was failure of the HPCI pump to operate.
This would result in anincrease in core damage frequency (CDF) of 3.66E-6/reactor year. The exposure time is estimated from whenthe last successful run of the HPCI Pump was performed on 5/23/13 at 0034 hrs until the HPCI System and flowcontroller was tested satisfactorily on 5/24/13 at 0230 hours. This results in approximately 26 hours of exposuretime and the incremental core damage probability (ICDP) is 5.43E-9, which is non-risk significant.
REPORTABILITY This event was reported to the USNRC via Event Report #49064 on 5/23/2013 pursuant to 10 CFR50.72(b)(3)(v)(B) and (D) -Any event or condition that at the time of discovery could have prevented fulfillment of the safety function of structures or systems that are needed to: Remove residual heat and Mitigate theconsequences of an accident.
NRC FORM 366A (10-2010)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
LICENSEE EVENT REPORT (LER)CONTINUA1. FACILITY NAMEPilgrim Nuclear Power StationTION SHEETPREVIOUS OCCURRENCES:
A review of Pilgrim Station License Event Reports (LERs) issued since year 2000 was performed.
The focus ofthe review was LERs that involved loss of HPCI system function or loss of system function due to flow controller malfunction.
The following LERs were reviewed:
LER 2000-002
-HPCI System Inoperable Due to Power Inverter Feed to Flow Controller Circuitry LER-2004-002
-HPCI System Inoperable Due to Fuse Failure in Gland Seal Condenser Circuit.LER 2004-004
-RCIC System Inoperable Due to Flow Controller Oxidation LER 2005-001
-HPCI System Inoperable Due to Fuse Failure in Motor Operated Valve Control CircuitLER 2008-004
-HPCI System Inoperable Due to Undervoltage Relay Failure in Valve Power Supply CircuitLER 2011-006
-HPCI System Inoperable Due to Governor Control Valve Mechanical BindingThese LER events do not identify any similar failure mechanisms to that described in this LER.In March 2012, Pilgrim Station identified defects in NUS Model PID901-540 flow controllers that were purchased to replace HPCI and RCIC System flow controllers (Condition Report CR-PNP-2012-1406).
A manufacturer report was generated to document the 10 CFR Part 21 Evaluation (No. 21-12-09).
The issue specifically addressed relates to flow controller setpoint thumbwheel manufacturing assembly defects.
Pilgrim sent the allflow controllers back to the manufacturer for reconditioning.
The condition addressed in this LER event report differs from the manufacturing defects evaluated in thevendor's Part 21 evaluation.
ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODESThe EIIS codes for Components and Systems referenced in this report are as follows:COMPONENTS CODESFlow Indicating Controller FICSYSTEMSHigh Pressure Coolant Injection (HPCI) BJ
 
==REFERENCES:==
 
Condition Report CR-PNP-2013-4286 and the associated Apparent Cause Evaluation Report; HPCI FlowController Failure to Achieve Rated Flow While in Auto.Condition Report CR-PNP-2012-1406, NUS Model 901-540 Flow controller defects.NRC FORM 366A (10-2010)}}

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