ML20217F232

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LER 970934-01:on 970918,RHR Pump Suction Relief Valve Setpoint Not IAW TS Was Determined.Caused by Insufficient Configuration Control.Declared B Train RHR Valve Inoperable, Recalibrated to Correct Setpoint & Declared Operable
ML20217F232
Person / Time
Site: Millstone Dominion icon.png
Issue date: 09/29/1997
From: Danni Smith
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20217F224 List:
References
LER-97-034, LER-97-34, NUDOCS 9710080022
Download: ML20217F232 (4)
v • d • e


Text

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NRcf oRM 366' U.S. NUCLEAR RLGuLAloRY CoMMibsloN Arraovto DV owe too stbo ono4 le 96) B AP'RE8 04!80'88 l '

'8 8.Ta % Y E l" I'"J M o" $ P M f " L'. 0 ,"" U N 8E l'

?(,f9;,'LWtba'&% t

'l"g.,M.',l?t',, n LICENSEE EVENT REPORT (LER)

(See reverse for required number of L mw=mmm,a'I"b'd'Nd g,'t 'cJ' lt,",*,','g.m,a na* eg-

, gag digits / characters for each block)

PCCaLifV hAME (1) DOCktiNUMetHth PAQt(M

Millstone Nuclear Power Station Unit 3 05000423 1 of 4 l

l 11f4E 64) l Residual Heat Removat Pump Suction Relief Valve Sotpoint Not in Accordance With Technical Specifications EVENT DATE (s) LLR NUMBER (6) REPORT DATE 17) o THER F ACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR stouENTIAL Revision MONTH DAY YEAR # Aclu 1Y NAME DOC &$ 1 NVMatM NUMBER NUMDER 09 18 90 96 - 034 - 01 09 29 97 l oPERATINo D THis REPORT IS SUDMITTED PURSUANT to THE REQUIREMENTSoF lu cFR 61 (CheCit one or more) (11)

I MODE (9) 20.2201(ts) 20.2203(aH2Hv) X; 60.73(aH2HI) bO.73(aH2Hvhi)

POWER. 000 20.2203(aH1) 20.2203(aH3Hil 60.73(aH2Hii) 60,73(aH2Hz)

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LEVEL (10) 20.2203(sH2Hil 20.2203(aH3Hn) bM3(aH2Hin) 73.71 20.2203(eH2Hn) 20.2203(aH4) 60.73(aH2Hiv) OTHER 20.2203(aH2 Haul 60.36(CH1) 60.73(aH2Hv) 20.2203(aH_2Hev) bO 36(cH2) 60.73(aH2Hvia) l gPgcg6n,Atget bei6w

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LICENSEE CONTACT FoR THIS LERi12)

NAMt TitlPHONE NUMatR sttictuce Ases Codel David A. Smith, MP3 Nuclear Licensing Manager (860)437 6840 COMPLETE ONE LINE FOR EACH COMPONENT FAILURF DESCltl8ED IN THis REPORT (13)

COUSE BY51EM COMPONt N T MANUf ACTURER HLPOH1 ABLE CAU$t $YSTEM COMPONt NT MANUF ACTURt R ht PQHI AHtt TO NPROS 10 NPRDS SUPPLEMENTAL REPORT EXPECTEDl14) EXPECTED MONTH DAY YEAR i X YES (if yes, complete EXPECTED sVBMISsioN DATL),

No suSMISSloN 11 22 97 DATE (15)

ADETRAct (L6mit to 1400 spaces,6.e., appromlmately16 single.spacedtypewnttenhnes) (16)

On September 18,1996 with the plant in Mode 5, during a review conducted by plant engineering pers0nnel, it was dCtermined that the actual setpoint for the Residual Heat Removal (RHR) pump suction relief valves was not in accordance with the requirements of Technical Specifications (TS). Contrary to the TS limit of 450 psig, the actuallift pressure for the RHR pump suction relief had been set at 440 psig. This condition was reported on October 18,1996 pursuant to 10CFR50.73(a)(2)(1)(B) as an operation or condition prohibited by the plant's TS, On August 29,1997 with the plant in Modo 5, while performing a review of RHR suction relief valve surveillance test procedures, it was discovered that previous RHR relief va've calibrations may have been performed improperly resulting in the B train RHR relief valve setpoint being outside the tolerance specified within the TS Limiting Condition for Operation. Consequently, this condition is being reported pursuant to 10CFR50.73(a)(2)(i)(B) as any operation or condition prohibited by the plant's TS.

The cause of the September 1996 event was attributed to insufficient configuration control within the design control program. The apparent cause of the August 29,1997, event was inadequate review of test procedures.

As an immediate corrective action for the August 29,1997 event, the B train RHR valve was declared inoperable, recalibrated to the correct setpoint and declared operable. Investigation into the cause of this event is ongoing. The RHR setpoint calculation will be revised to properly incorporate the RHR pump suction relief valve temperature correction for the full temperature range. An evaluation of the RHR relief valve setpoint effect on the RHR piping stress analyses will be performed.

9710090022 970929 PDR ADOCK 05000423; 6 PM l

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NRc f ORM 366A U.s. NUCLEAR REluLAToRY Commission i LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION P ACILITY NAME 11) DOCKET NUMBER (2) LER NVMSER ts) PAoE 13)

YEAR SEQUENTIAL Revision Millstona Nuclear Power Station Unit 3 05000423 NUMBE1 NUMBE R 2 of 4 96 - 034 -

01 1LXT lit more space is newted, vse ed&tionalcopies of NRC form 366A) l1 n

1. Descriptlon of Event On September 18,1990 with the plant in Mode 5, during a review conducted by plant engineering personnel, it was det:rmined that the actual setpoint for the Residual Heat Removal (RHRt pump suction relief valves was net in accordance with the requirements of Technical Specii'ication 3 4.9.3 a.2 and 3.4.9.3 a 3. Technical Specifications (TS) require that the RHR pump suction relief be set at 450 pounds per square inch gauge (psig)in order to provide adequate over pressure p'rotection when the temperature of any Reactor Coolant System (RCS) cold leg is less than

< 350 degrees Fahrenheit ( F). Contrary to this requirement the actuallift pressure for the RHR pump suction relief had been set at 440 psig. This condition was reported on October 18,1990 pursuant to 10CFR50.73(a)(2)(i)(B) as an operation or condition prohibited by the plant's Technical Specifications.

On July 10,1997. License Amendment 143 was incort wated resulting in an RHR relief valve satpoint of 2 426.8 psig and s 453.2 psig. On August 29,1997, with the pl% in Mode 5, while performing a review of RHR suction relief valve surveillance test procedures, it was discovered that previous RHR relief valve calibrations may have been performed improperl,. This condition was not identified during reviews performed in conjunction with implementation of the license amendment.

Th3 valve vendor requires a setpoint temperature correction factor of +3% when valves are used in 250' F or higher r,cryice and tested / set at ambient temperatures. This is to compensate for changes in spring loading. At MP3, the valves have always been tested and calibrated in this manner. At cold conditions (<250* F), the vendor's +3%

t:mperature correction factor no longer applies. Since this relief valve must serve in both temperature ranges, this 3 cffectively narrows the setpoint acceptance tolerance to 0%, +3% or (440 to 453.2 psig). The actual B train RHR relief valve setpoint was 455 psig and not within the Limiting Condition for Operation (LCO) tolerance. Consequently, this condition is being reported pursuant to 10CFR50.73(a)(2)p)(B) as any operation or condition prohibited by the plant's Technical Specifications.

II, Cause of Event On March 26,1985 during the construction of the unit, the Architect / Engineer (A/E) issued a design change, which lowered the setpoint for the RHR suction relief valve from 450 psig to 440 ps:g. This design change did not specify that a technical specification change was required. The design change also resulted in the issuance of a setpoint calculation change on April 29,1985 which confirmed the change from 450 psig to 440 psig. Subsequently, the plant M:intenance Department established the relief valve test program in accordance V,lth the requirements imposed by American Society Of Mechanical Engineers (ASME). In implementing the relief valve test program, reliance was placed on the information contained in setpoint calculations and design changes when establishing the appropriate values for t; sting of valves contained within the program.

The apparent cause of the August 29,1097, event involving the identification of incorrect previous RHR suction relief valve calibrations is inadequate review of test procedures. This condition was not identified during the reviews performed in conjunction with implementation of a TS license amendment to revise the RHR relief valve setpoint.

NRC FORM 366A 44-95)

NRc f ORM 366A U.s. NUCLEAR RE LUL AToHY COMMISSION UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION 7 AclLITV NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAoE (3)

YLAH SLoulNTIAL ftLVISloN Millstone Nuclear Power Station Unit 3 05000423 wtaer 1 NUMBER 3 of 4 90 -

034 -

01 TLKT (11more space os required use additionnicopies of NRC form 366M (11) 111. Analysis of Event The unit has operated using the 440 psig setpoint as required by the original A/E design change and the associated setpoint calculation. The 440 psig setpoint is the fechnically correct setpoint for the RHR suction relief valves to ensure cv;r pressure protection for the RHR piping in accordance with /sSME requirements. Subsequent relief valve capacity calculations were based on the 440 psig setpoint. Over pressure protection of RHR piping and Low Temperature Over j pressure / Cold Over Pressure Protection System (LTOP/COPPS) requirements also f at i utilized the 440 psig setpoint for calculations performed following the initial design change.

Fcr the Augus' 1997 event, the ability of the RHR relief valves to function as a Cold Over Pressure Protection (COPPs) r:12f valve % demonstrated in an engineering calculation. With the relief valves set to the TS values the maximum I

reactor csolant system pressure is $47 psig which is below the Appendix G overpressure limit of 558 psig.

Consequently, the maximum pressure limit of the reactor vessel was not challenged as a result of the TS setpoint being exceeded by 2 psig. Preliminary evaluation indicates that the pipe stresses within the RHR System are acceptable.

This report will be supplemented based on the results of the final evaluation.

Based on the preliminary evaluation, the over pressure protection functions were not compromised as a result of the reIZf valve exceeding the TS setpoint, the safety significance was minimal and there were no adverse safety consequences as a result of the event.

IV. Corrective Action As a result of the September 18,1996 event, a Technical Specifications Change was implemented on July 10,1997 that revised the setpoint value identified in Technical Specifications 3.4 9.3.a 2 and 3.4.9.3 a.3 from 450 psig to 2 426.8 psig and s 453.2 psig.

As an immediate corrective action for the August 29,1997 event, the B train HHR was declared inoperable. After correcting the surveillance testing procedure to take into account the valve's hot / cold settings (-0%, +3%) and gauge accuracies the valve was calibrated to the correct setpoint.

The following corrective actions will be performed:

1. Investigation into the cause of this event is ongoing. A supplement to this report will be submitted based on the results of this investigation prior to entry into Mode 4.
2. The setpoint calculation will be revised to properly incorporate the RHR pump suction relief valve temperature correction for the full temperature range prior to entry into Mode 4.

3, An evaluation of the RHR relief valve setpoint effect on the RHR piping stress analyses will be performed prior to entry into Mode 4.

V. Additional Information None NRC FOS 360A (4-95)

wnc fonwi[shA U.s. NUCLEAR REIULATcRY Commission LICENSEE EVENT REPORT (LER) )

TEXT CONTINUATION l FAclLITY NAME 11) DOCKET NUMsER (2) LER NUMBER (6) PAGE (3)

YEAR SEQUENilAL RLvlSION Millstone Nuclear Power Station Unit 3 05000423 NUMBH NUMBER 4 of 4 96 - 034 -

01 ,

TEKT illmore spaceis required. use addalonalcopies of NRC /orm .166As t11)

Similar Events Listed below are historical LERs whose causes can be attributed to insufficient configuration control within the design control program. Many of these events were identified as a result of the Configuration Management Review Process.

LER 96-007-00 Containment Recirculation Spray, Quench Spray, and Safety injection System Outside Design Basis Due to Design Errors LER 96-009-02 Inoperable Shutdown Margin Monitors from Low Count Rate, Due to inadequate Design Control LER 06-013-00 Residual Heat Removal System Design Deficiency Due to Non-conservative Original Design Assumption LER 96-026-02 Non-Conservative Primary Grade Water Flow Rates Used in Boron Dilution Safety Analysis LER 97-003 00- Potential For Recirculation Spray System (RSS) Piping Failure Due To RSS Pump Stopping And Restarting During Accident Conditions LER 97-015-00 Potential Vortexing of Recirculation Spray System Pumps LER 97-02100 Defective Design of RSS Expansion Joint Tie Rod Assembly LER 97 028-00 PotentialLoss of Net Positive Suction Head for Recirculation Spray System Pumps LER 97-029-00 Design Basis Concern on SGTR Analysis for MSPRBV .

LER 97-03100 RHR Valve Low Pressure Open Permissive Bistable Setting Set Non-Conservatively LER 97 035-00 Potential Nonconservatismfor Steam GeneratorWater Level Low Low Trip Setpoint Due to PMA Term Uncertainties LER 97 046-00 Containment Recirculation Spray System Cubicle Flood Potential Manuf acturer Data Ells System Code R Isid ual H e at Removal (PWR) . .... .. ... .. .. . . ... . .. . . ....... . . .... ... ... ..B P 4

Ells Component Code V ci v e , R e li e f . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . R V I

l NGC FORM 366A 14 9H -

4 _ , , , - _ . . . . _ , _ _ _ , - . . . . . _ .

_ _ _ , _ _ _ _ . _ , . . . _ . . . - , _ _ - _ . _ _

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