ML18101B200

LER 95-027-00:on 761211,doses During LOCA Exceeded Plant Licensing Basis Due to Inaccurate Assumptions in Dose Calculations.Revised Procedures in August 1994.W/forwarding Ltr

ML18101B200
Person / Time
Site:	Salem
Issue date:	11/27/1995
From:	SHUMAKER D, WARREN C C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-95-027-01, LER-95-27-1, LR-N95207, NUDOCS 9602070146
Download: ML18101B200 (6)

Text

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• Public SeNice Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit NOV 2 4 1995 LR-N95207 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

LICENSEE EVENT REPORT 272/95-027-00 SALEM GENERATING STATION -UNIT 1 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 This Licensee Event Report entitled "Operation of Positive Displacement Pump During a Safety Injection Could Have Resulted in Exceeding lOCFRlOO and GDC 19 Dose Criteria" is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR50. 73 (a) (2) (i) . SORC Mtg.95-140 Attachment DVH/tcp C Distribution LER File 9602070146 951127 PDR ADOCK 05000272 S PDR Sincerely, @A-Cla'?:lc.

Warren General Manager -Salem Operations The power is in yDUr hands. 95-2168 REV. 6 94

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• Attachment A PSE&G Commitments for LER 272/95-027 The following items represent PSE&G commitments made to the Nuclear Regulatory Commission related to LER 272/95-027-00.

The commitments are as follows: Perform a comprehensive review of the UFSAR Chapter 15 accident analysis to better ensure the consistency between the design assumptions, and plant configuration or operation.

Identify and document all sources of contaminated leakage, including a secured PDP, into the Auxiliary Building during a LOCA and incorporate in dose calculations.

Review open items identified in the ABV system CBDs to ensure that all issues that could impact the plant licensing basis have been addressed to ensure no safety concerns exist post restart. Also, identify improvements to the ABV system and/or revisions to the EOPs necessary to ensure compliance with lOCFRlOO limit and GDC 19 limit. Incorporate changes into the impacted documentation, including the UFSAR. The above items will be completed prior to restart. -

• NRC FORM 366 U.S. IU:LEAR REGIJLATORY aHllSSUlll APPROVED BY Dm m. 3150-0104 (4*95> . EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.

FORWARD (See for required ni.iiber of CC>>4MENTS REGARDING BURDEN ESTIMATE TO THE reverse INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 digits/characters for each block) U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON D 20555-0001, AND TO THE PAPERWORK REDUCTION FACILITY 11A1E (1) DOCKET IUllER (2) PAGE (3) SAUM GENERATING STATION UNIT 1 05000272 1 OF 4 TITLE (4) Operation of Positive

!\1ItP rurµg a Safety Injection could have Resulted in lOCFRl.00 am 19 nose Limit Criteria EVENT DATE f5> LEI lllllER f6 REPCJIT DATE f7> OTHER FACILITIES llMJLVED fB> YEAR I SEQUENTIAL REVISI FACILITY NAME DOCKET NUMBER MONTH DAY YEAR ON MONTH DAY YEAR NUMBER NUMBER SAUM UNIT 2 05000311 FACILITY NAME DOCKET NUMBER 12 11 76 95 -027 -00 11 27 95 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE RECIJIREMENTS OF 10 CFR §: (Check one or more ctn MOOE (9)

• 20.2201(b) 20.2203(a)(2)(V) x 50. 73(a)(2)( i) 50. 73(a)(2)(vi ii POER 20.2203(a)(1) 20.2203(a)(3)(i)

50. 73(a)(2)( ii) 50. 73(a)(2)(x)

LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 illl!lllllllllll1l1'll1ll 20.2203(a)(2)(ii) 20.2203(a)(4)

50. 73(a)(2)( iv) OTHER 20.2203(a)(2)Ciii) 50.36(c)(1)

50. 73(a)(2)(v) in Abstract 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii) below or in NRC Form 366A LICENSEE CDITACT FOR THIS LER 12) NAME TELEPHONE NUMBER (Include Area Code) Mr. D. Shtnnaker, Specialist EngineerinJ SUpervisor 609-339-1852 CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE ll1itl 1 lif 11111 CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NP RDS TO NPRDS I SUPPLEMENTAL REPORT EXPECTED C14> EXPECTED MONTH DAY YEAR 'YES XINO SUIMI SS Ull (If yes, EXPECTED SUBMISSION DATE). DATE (15) ABSTRACT CL imit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) On October 26, 1995 it was detennined that doses durinJ a IOCA could have exceeded the plant licensirg basis due to inacx::urate assunptions in the dose calculations.

'!he previous acx:::ident analysis was perf onned assunrin:J the PDP trips after an safety injection signal am that the exhaust fran the Auxiliacy Buildirg passes through a dlarcoal filter durirg the cold leg recirculation of a IOCA. '!he positive displacement

?IIIP in the Cllemical am Volmne Control system only trips corx::urrent with a loss of offsite pc:Mer after a safety injection signal. In addition, the Auxiliaey Buildirg Ventilation dlarcoal filtration is not autanatically aligned in acx:::ident c::omitions.

Withalt the dlarcoal filtration, the calculated site doses due to leakage of the PDP seals durirg a IOCA are significantly greater than the doses in the UFSAR ani may exceed the lOCFRl.00 site boun:lary dose limits arrl the GOC 19 control roam limits. 'lhe Energency q>eratirg Procedures were r:evised in August 1994 to manually trip the PDP prior to enterirg cold leg recirculation.

'!his event is reportable in accordance with lOCFR 73(a) (2) (ii) I any event that resulted in a comition that was outside the design basis of the plant. NRC FORM 366 (4-95)

• U.S. llJClEAI REGULATORY COIUSSION LICENSEE EVENT REPORT (LER) TEXT CDNl'INUATION FACILITY NAME (1> DOCKET LER 11.11BER (6) PAGE Cl) YEAR I SEQUENTIAL I REV NUMBER NUMBER 2 OF 4 05000272 SAIBM GENERATING S'm'l'ION UNIT 1 95 -027 -00 TEXT (If more space is required.

use additional copies of NRC Form 366A) (17) PIAN'!' AND SYSTEM IDENI'IFICATION Westin;Jhouse

-Pressurized Water Reactor Positive Displacement Pl.mp , Cllemical

& Volume Control System (OTC) {CB/P}*

• Energy :rmust.ry Identification System (EIIS) codes am caaponent furx::tion identifier codes Cl£P?ar in the text as {SS/CT:C}.

IDENI'IFICATION OF OCXIJRRENCE EVent rate: Unit 1: Decanber 11, 1976 (Initial Plant Critically)

Unit 2: August 2, 1980 (Initial Plant Critically) rate Determined to be Reportable:

Oct:cber 26, 1995 <XlNDITIONS PRIOR 'ID OCXl.JRRENCE unit 1: unit 2: Defueled, 000 % Reactor Power M:Jde 5, 000 % Reactor Power '!here -were no structures, caiponents, or systems that were inoperable at the start of the event that caitrib.rt:.ed to the event. DESOUPr!ON OF OCCURRENCE On Oct:cber 26, 1995 it was determined that doses durinJ a Loss Of Coolant Aa::ident (I.OCA) oculd potentially exceed the lOCFRlOO site bourrlacy dose limit am GOC 19 control roan dose limit. Previa.is analyses assrmv;,d that the positive displacement pmp (PDP) in the Olemical am Volume Control (OTC) system tripped after a safety injection (SI) signal. However, it was determined that the PDP only trips concurrent with a loss of offsite J;XJWer after a safety injection signal. DlrIDJ the cold leg recirculation i;ilase of a I.OCA, q;>eration of the PDP results in leakage of contaminated water past the PDP seals. '!he PDP seal leakage increases the total contaminated leakage to the Auxiliary BuildinJ.

'!he UFSAR accident analyses were perfonned assurniig the PDP trips after an safety injection signal am do not consider contaminated leakage from the PDP seals. DlrIDJ a I.OCA, the charcoal.

filter for the Auxiliary Buildin;J Ventilation (ABV) system is not autana.tically aligned but nust be aligned by an operator fram the control roan. In the UFSAR accident analysis, it is assumed that the exhaust from the ABV system passes through the charcoal filter durinJ the recirculation

Eilase of the I.OCA. Without the dlarcoal.

filtration, calculated doses durinJ a I.OCA are significantly greater with PDP seal leakage am may exceed the lOCFRlOO site bourrlacy dose limits am the G1X! 19 control roam limit. NRC FORM 366A (4-'I:>

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• U.S. llJClEAR IE&Jl..ATmt1' aJllJISSllll LICENSEE EVENT REPORT (LER) 'l'E}CI' CDN.l'INUATION FACILITY' NAME (1) DOCKET LER IUIBER (6) YEAR I SEQUENTIAL I REVIS! NllMBER ON SAUM GENERATING STATION UNIT* 1 05000272 i 95 .. 027 .. 00 TEXT (If more space is required, use additional copies of NRC Form 366A> (17> ANALYSIS OF OCXl.JRRENCE PAGE (3) 3 OF 4 '!he IDP is part of the eve system am is used for nonnal makeup, boratian am dilution c:perations.

'1he PDP is located in the Auxiliai:y Build.irg.

In addition to the PDP, there are two dlargmysafety injection (SI) p.mp; in the eve system. rurin} oonnal operation, these pmps provide additional makeup capacity when the system demani exceeds the capacity of the PDP. On receipt of a safety injection signal, both SI :puIIpS start autanatically.

PDP c:peration is not required for a safety injection.

Followi.n;J the April 7, 1994 Reactor Trip am Safety Injection event, review of the event identified that the PDP was operatin}

prior to the SI am that the PDP did not trip after receipt of the SI signal. Review of the instrumentation am control drawin.;Js oonfinned that the PDP control system was arrarged such that the PDP will not trip on receipt of a SI signal. An evaluation to detennine the effect of the PDP operation durin} a SI was oarpleted by Westin:;Jhouse in August 1994. '!he PDP seal leakage durin} punp operation can significantly increase the site bamdary am control roan dose. '!he Westin;Jhouse evaluation concluded that the increased dose due to PDP seal leakage durin} a IOCA would still meet GDC 19 am 10 CFR 100 requirements as lcn;J as the IDP seal leakage does not exceed 2 gpn. '1he Procedures (EDPs) TNere revised in August 1994 to manually trip the IDP prior to enterin} cold leg recirculation.

On Oct:d:>er 26, 1995, durirg evaluation of an autanatic trip function for the PDP, the issue of the PDP conti.nuirg to operate after a SI signal am its consequerx::es were reviewed for reportability.

rurin} this review, the Westi.Djlouse dose evaluation was detennined to be in error in that it assumed the A'fN system charcoal filter was aligned to provide filtration duri.rg the cold leg recirculation piase of a I.DCA. '!he A'fN system dlarooal filter is not autanatically aligned. Without charcoal filtration, calculated doses increase significantly. when the filter is aligned, the doses may exceed the 10 CFR 100 am the GDC 19 limits. APPARENI' CAUSE OF OCCIJRRENCE

'!he design basis for the operation am control of the PDP is not clearly dcx:::l.llnente in the UFSAR. '1he UFASR does not di saJSS the need to trip the PDP prior to cold leg recirculation durin} a IOCA. '1he incatplete doCl.nnentation of the eve system resulted in plant c:peration in a configuration which was not considered in the UFSAR accident analyses.

'!he desicjn basis documentation for the MN system is not canplete.

'1he :requirement to align the c:harooal filter prior'to cold leg recirculation duri.rg a I.DCA is not specified .iJi the UFSAR. . mIOR SIMIIAR CXXIJRRENCES lll.ere were two previous IERs where the IDP operation was a factor. I.ER 272/94-007-00 addressed the April 7, 1994 Reactor Trip arrl Safety Injection, am 272/94-017-01 addressed Pressurizer overpressure Protection bein;J outside of design basis. NRC FORM 366A (4*95))

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• lllC FCllllll 366A *

-U.S. llJCLEAR REGUl.ATCllY CXMUSSIOll L:ICENSEE EVENT REPORT (LER) TfilIT CXNI'INUATION FACILITY """" (1) DOCKET LER IUllER (6) YEAR I SEQUENTIAL I REVIS[ __ NUMBER ON 95 *-027 .. 00 05000272 SAUM. GENERA'l:U:.G STATION UNIT 1 TEXT (If more space is required, use additional copies of NRC Form 366A> (17) RUOR SIMIIAR OCCIJRRENC&S (cont'd) PAGE (3) 4 OF 4 IER 272/95-017 identified that the operation of the Eme1:'gency Air ccn:li.tionin:;J System (EACS) as it was configured

'WOUld have :resulted in GOC 19 criteria for control roan habitability bei.rg exceeded.

Awareness of this I.ER aided in the identification of the PDP operation possibly exceeding the lOCFRlOO am GOC dose limits. '!\<<> IERs in 1995 identified incidents

'Where the UFSAR am the ac:x::ident analysis were in conflict.

I.ER 272/95-011-00 identified that a discrepancy existed between the UFSAR am the steam line break analysis for the oontairnnent penetration area. I.ER 272/95-016-00 identified a discrepancy

involvi.rg the oontairnnent structure duri.rg a steam Line Break Analysis ani a Design Basis I.DCA. Both IERs are similar to this I.ER in that a discrepancy existed between the ac:x::ident analysis am the UFSAR. SAFEIY SIGNIFICANCE

'1he operation of the PDP after a lOCA has a direct i.npact on the potential for off-site release of radioactivity.

'lhe deficiency in the design basis could have resulted in site bcumary ani control roan doses exceeding licen.si.rg basis limits duri.rg a lOCA at either Salem Unit. A review of the Westin:]house evaluation revealed that an offsite dose in excess of the lOCFRlOO am GOC 19 limits ooold have ocx:urred a.ssumin;J a IDCA am an mifiltered release (dlarcoal filter not on-line).

'!he design basis sairce tenn values are considered to be conservatively high. Based on the sam:e tenn definition aooepted for the Mvanoed Light water Reactor design, the NRC ani the in:iustry are evaluatin_J the use of ioore :realistic source tenn predictions.

Use of the m:>re :realistic source tenn values in analyzi.rg the operation of the PDP during an SI inlicate that the 10 CFRlOO ani GOC 19 criteria would not likely have been exceeded if an event had occurred.

1. '1he EDPs were revised to include steps for manually triwi.rg the PDP prior to ocmnerx::i.ng cold leg recirculation.

2. Perfonn a ccmprehensive review of the UFSAR ClJapter 15 ac:x::ident analysis to better ensure the consistency between the design assumptions, am plant configuration am operation.

3. Identify am d<x:ument all sources of contaminated leakage, incluclir¥]

a secured PDP, into the .Auxiliary Building.duri.ng a IDCA am incorporate*

in dose calculations.

4. Review open items identified in the MN system CBDs to ensure that all issues that could inpa.ct the plant licensing basis have been addressed to ensure no safety concems exist post :restart.

Also, identify inprovements to the MN system aOO,lor revisions to the EDPs necessary to ensure corrplianc:e with lOCFRlOO limit am GOC 19 limit. Incorporate changes into the i.npacted docurmmtation, inch.r:li.rg the UFSAR. Items 2-4 will be ccmpleted prior to restart. NRC FORM 366A (4*95) -. ---;* :-..