ML17285B347

LER 89-040-01:on 890919,determined That Under Certain Meteorological Conditions Situation Would Be Created Not within Licensing Basis Consideration for Secondary Containment performance.W/900619 Ltr

ML17285B347
Person / Time
Site:	Columbia
Issue date:	06/19/1990
From:	Arbuckle J, Powers C WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-040, LER-89-40, NUDOCS 9006270256
Download: ML17285B347 (11)

Text

ACCELERATED D STRIBUTION DEMONS TION SYSTEM 1 4 REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR: 9006270256 DOC. DATE: 90/06/19 NOTARIZED: NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION ARBUCKLE,J.D. Washington Public Power Supply System POWERS,C.M. Washington Public Power Supply System RECIP.NAME . RECIPIENT AFFILIATION

SUBJECT:

LER 89-040-01':on 890919,standby gas treatment sys capability not within license basis consideration.

W/9 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

NOTES'ECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL

'

ID CODE/NAME LTTR ENCL PD5 LA 1 PD5 PD 1 1 SAMWORTH,R 1 1 INTERNAL: ACNW 2 2 ACRS 2 2 AEOD/DOA 1 1 AEOD/DS P/TPAB 1 1 AEOD/ROAB/ DS P 2 2 DEDRO 1 1 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB9H3 1 1 NRR/DLPQ/LHFB11 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB11 1 1 NRR/DREP/PRPB11 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR~Sg/:S PEED':~ 1 1 NRR/DST/SRXB 8E 1 1 1 1 RES/DSIR/EIB 1 1 RGN5 FILE 01 1 1 EXTERNAL EGGG STUART i V A 4 4 L ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 -

1 NSIC MAYS,G 1 1 NSIC MURPHY,G.A 1 1 NUDOCS FULL TXT 1 1 rv~

t oust'o 2<~9 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 36 ENCL 36

WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 96B ~ 3000 George Washington Way ~ Richland, Washington 99352 Docket No. 50-397 June 19, 1990 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

NUCLEAR PLANT NO. 2 LICENSEE EVENT REPORT NO. 89-040-01

Dear Sir:

Transmitted herewith is Licensee Event Report No. 89-040-01 for the WNP-2 Plant. This report is submitted in response to the report requirements of 10CFR50.73 and discusses the items of reportability, corrective action taken, and action taken to preclude recurrence.

'ery,t.ruly yours, gttt/.,:.,

C. H. Powers (H/D 927H)

WNP-2 Plant Hanager CHP:lr

Enclosure:

Licensee Event Report No. 89-040-01 cc: Hr. John B. Hartin, NRC Region V Hr. C. J. Bosted, NRC Site (H/D 901A)

INPO Records Center Atlanta, GA Hs. Dottie Sherman, ANI Hr. D. L. Williams, BPA (H/D 399)

NRC FORM 366 , U.S. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 3(504)104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE To THE RECORDS AND REPORTS MANAGEMENT BRANCH IP4I30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PA E 3 Washin ton Nuclear Plant - Unit 2

'"" '" Standby Gas Treatment ystem 0 s 0 o 03 97 ior0 8 apa i i y o

-

i in i cense asi s onsi era i on for Secondary Containment Performance Under Certain Conditions Due to Design EVENT DATE (6) LER NUMBER (6) REPORT DATE I7) OTHER FACILITIES INVOLVED (SI MONTH DAY YEAR i%+ SEOUENZrAL REYrSK FACILITYNAMES DOCKET NUMBER(SI YEAR ?K~4 NUMSER gg NUMSER MoNTH OAY YEAR 0 5 0 0 0 0 919 8 9 8 9 040 01 06 19 9 0 0 5 0 0 0 OPERATING THIS REPORT IS SUBMITTED PURSUANT To THE RLOUIREMENTS OF 10 CFR (): /Check one or morr ot the /or/or>>inc/ (ill MOOE (9) 20.402(5) 20A05(cl 60.73( ~ l(2)liv) 73.71(tr)

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LICENSEE CON1'AC'T FOR THIS LER (12)

NAME TELEPHONE NUMBER AREA CODE J . D. Arbuckle Com liance En ineer 50 937 7- 211 5 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPOR'7 (13)

CAUSE SYSTEM COMPONENT MANUFAC REPORTABLE MANUFAC EPORTABLE gS Sr CAUSE SYSTEM COMPONENT TURER TO NPRDS

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On September 19, 1989 it was determined by Engineering analysis that under certain meteorological conditions (moderate wind and low temperature), coincident with a DBA LOCA and assumed failure of one train of the Standby Gas Treatment (SGT) system, a situation would be created that is not within the licensing basis consideration for Secondary Containment performance. The Engineering analysis was performed as a further corrective action for LER 88-023.

The QNP-2 FSAR states that the Secohdary Containment will be maintained at=minimum differential pressure of -0.25H W.G. following a postulated LOCA, and that this differential will be established within two minutes following the accident. Recent analysis, based upon Standby Gas Treatment, Secondary Containment, Standby Service Water and weather modeling, shows that during post-LOCA, or adverse weather, differential pressure of the'Secondary Containment may not always meet the FSAR commitments. Certain combinations of post-LOCA single active failures and winter conditions adversely affect Secondary Containment and, as a result, increase Secondary Containment leakage.

NRC Form 366 (64)9)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (669) APPROVED OMB NO. 31500)04 EXPIRES: 4/30/92 4ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EV REPORT (LER) INFORMATION COLLECTION REQUEST: 60A) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P.530), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31504104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON,DC20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

LER NUMBER (6)

YEAR ~g: SEOUENTIAL REVISION NUMBER NUMBER Mashin ton Nuclear Plant TEXT /I/ moro 4/>>ce /4 rer/oired, ore eddie'ooe///RC %%dnrr

- Unit 366A'4/ () 7) 2 o s << o 9 7 9 0 0 4 0 010 2 oFO 8 On. January 8, 1990, as a result of in-depth reviews of calculations pertinent to secondary containment accident performance, a problem with the meteorology data presented in FSAR Amendment 36 was discovered. It was determined that the values for the source term dispersion pattern relative concentration factor, X/g, were specified incorrectly in FSAR Amendment 36.

As an immediate corrective action, a Justification for Continued Operation (JCO) was performed and concluded that operation of the Plant can continue while final resolution of this issue is achieved. In addition, this. situation was reviewed relative to the requirements of 10CFR50.59 and it was determined that it represents an unreviewed safety question. Accordingly, the NRC was formally notified of this determination.

As a further corrective action, a test was run to confirm the leakage value used for the JCO. In addition, Design Basis changes will be evaluated to provide an SGT system that allows for adequate filtering of Secondary Containment for applicable meteorological conditions, and system draw down time, so as to meet 10CFR100 and GDC 1 9 limits, while taking credit for suppression pool scrubbing as allowed by Standard Review Plan 6.5.5.

This event did not affect the health and safety of either the public or Plant personnel.

Plant Conditions Power Level - lOOX Plant Mode- - 1 (Power Operation)

Event Descri tion On September 19, 1989, it was determined by Engineering analysis that under certain meteorological conditions (moderate wind and low temperature), coincident with a DBA LOCA and assumed failure of one train of the Standby Gas Treatment (SGT) System, a situation would be created that is not within the licensing basis consideration for Secondary Containment performance. The Engineering analysis was performed as a further corrective action for LER 88-023, HTechnical Specifiation Violation of Secondary Containment to Outside Differential Pressure Caused by Design due to Programmatic Errors."

NRC Form 366A (64)9)

NAC FORM 366A U.S. NUCLEAR AEGULATOAYCOMMISSION (6$ 9I APPROVED 0MB NO. 31500106 E XP I R ES: 4(30(92 MATEO BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE REPORT (LER) E INFORMATION COLLECTION REOUESTI 50AI HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P430L U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500(041. OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1( DOCKET NUMBER (2( LEA NUMBER (6( PAGE (3I YEAR ~(@ SEOUENTIAL NUM SEA REVISION NUMBER Washin ton Nuclear Plant TEXT Irfmoro spooo is ror(rrirrrd

- Unit 0>> oddirlonsl NRC Form 366ABI (17(

2 o s << o 3 9 7 9 0 0 4 0010 3 oF 0= 8 The MNP-2 FSAR states that. the Secondary Containment will be maintained at minimum differential pressure of -0.25H W.G. following a postulated LOCA, and that this differential will be established within two minutes following the accident. Recent analysis, based upon Standby Gas Treatment, Secondary Containment, Standby Service Water (SSW), and weather modeling, shows that during post-LOCA, or adverse weather, differential pressure of the Secondary Containment may not always meet the FSAR com-mitments. Certain combinations of post-LOCA single active failures and winter conditions adversely affect Secondary Containment and, as a result, increase Secondary Containment leakage.

The analysis uses the lowest monthly average temperature for January of 12/F in combination with the highest average monthly wind for January of 1 0.3 mph. On the average, temperature is below 12/F approximately 1.6% of the calendar year, and below 0/F approximately 0.1% of the calendar year. Wind conditions above 10.3 mph will probably provide sufficient dispersion to preclude the need far maintaining the

-0.25H differential and; therefore, negates designing the SGT for worst case wind conditions.

Mind increases the demand on the SGT to hold the leeward side and roof of the Reactor Building sufficiently negative while simultaneously increasing the differential pressure and, thus, the inleakage on the windward side of the building. Differen- tial temperature between the inside and outside of the building creates a differential pressure gradient from the bottom to the top of the Secondary Containment due to the density difference of the air inside and outside the building during cold outside conditions. As a result, the lower portion of the building must be held at a high differential pressure (up to -0.75H) to assure that a -0.25R differential exists at the building roofline. This overall greater differential pressure proportionally increases building inleakage. The effects of wind and winter temperatures result =in the inability to hold the upper portion of the Secondary Containment at a -0.25H differential in cold and mildly windy weather, and lengthens the time required to reach -0.25H differential in warmer and less wing weather.

Analysis shows that the time required to reach the steady state condition is a function of the assumed meteorological conditions at the time of a postulated LOCA, type of single active failure coincident with the LOCA, and the Standby Service Water (SSW) temperature. The transient analysis clearly indicates that the limiting single active failure is the assumed loss of one SGT train. Based upon single train design basis SGT flow and maximum Technical Specification allowable Secondary Con-tainment leakage, the uppermost inside surface areas of the Reactor Building cannot be maintained at a -0.25 '.G. with respect to atmospheric pressure during low temperature and high wind conditions. High SSW water temperature acts to extend the time required to reach a steady state condition, but does not effect the final steady state differential pressure.

NRC Form 366A (6J(9(

NRC FORM 366A U.S, NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 31500104 EXP IR ES: 4i30i92 rMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EV REPORT (LER) E INFORMATION COLI.ECTION REOVEST: 60A) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION ANO REPORTS MANAGEMENT BRANCH IP.530), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (11 DOCKET NUMBER (21 LER NUMBER (6) PAGE (3)

YEAR @8 SEOVSNTIAL NUMBER REVISION NVM ER Washin ton Nuclear Plant - Unit 2 97 90 0 4 0 OF TEXT (Ji moro spssoir rsr)oirod, oss sddidonsl HRC Form 366A'sl (17)

With two fans redundantly powered in each train, the SGT is not susceptible to many of the single active failures that have a higher probability of occurrence relative to other events, e.g., failure of an emergency diesel generator to start. If one train does fail to start automatically, remote manual initiation and process monitoring can occur through the control room. A design review of the system to determine the susceptibility of an SGT train to single failure has not been performed. Until that occurs, the likelihood of failure, or what would be necessary to remedy failure susceptibilities, is not known. (Local control is not possible due to the post-LOCA radiation fields that are postulated to be present in the vicinity of the SGT trains.) From a failure analysis perspective, the SGT train design at WNP-2 does have features that provide more reliable operation than are dictated by the minimum design requirements that allow for satisfying single failure criterion by the existence of a redundant train.

Testing conducted during the past calendar year of SGT flow/differential pressure capability, and testing of Secondary Containment integrity show that the SGT is capable of performance beyond design basis requirements, and that the Secondary Containment is significantly more leak-tight than required by Technical Specifications. Actions have been taken over the past twelve months to further tighten the Secondary Containment boundary against leakage, e.g., Reactor Building Exhaust and Outside Air (REA and ROA) isolation valve seals have been replaced and the railroad bay door seals have been adjusted. Reanalysis using documented actual performance values for SGT flow capability and Secondary Containment leakage shows that post-LOCA pressure stabilizes at -0.32R with an outside temperature of 12/F with a coincident 10.3 mph wind, which is well below the required -0.25". However, the -0.25" level is not reached for approximately 3.5 minutes after the accident.

Additional margin to the design basis requirements is also available from the actual leakage performance of the Primary Containment. Table 1 outlines the results of analysis based upon licensing basis SGT and Secondary Containment performance fol-lowed by reanalysis results based on realistic SGT and Secondary Containment performance.

Table 1 also demonstrates that the plant can be maintained at the required negative pressures (albeit the time is greater than two minutes) with the current leak-tightness of the Secondary Containment and SGT capability at very low winter temperatures, i .e., -8/F with a 1 0 mph wind, and -23/F without wind. This i s obtained provided that the leak-tightness of Secondary Containment and/or the flow capability of SGT do not degrade by more than 5%, a differential of -0.25R can be maintained at 12/F with a 10 mph wind. Requirements for residence time in the SGT charcoal filters is met with at the 5600 cfm flowrate for design basis active and passive failure scenarios.

Provided that the SGT set point pressure is sufficiently negative, the existing SGT pressure control loop instrumentation will assure that the SGT trains operate at 5600 cfm flow as required during all meteorological conditions. Exisitng loop instrumentation controls Secondary Containment pressure during windy conditions up to existing REA or SGT capacity.

NRC Form 366A (64)9)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NO. 31504)104 EXPIRES: 4/30/92 LICENSEE EV ,4ATED BURDEN PER RESPONSE TO COMPLY WTH THIS REPORT (LER) INFOAMATION COLLECTION REQUESTI 60A) HAS. FORWARD COMMENTS REGARDING BUADEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P4301, U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31604)104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR < SEOVSNTIAL gag( REVISION

~~<? NVMSSR ~ OS NUM 8 SR Plant - U i 0 5 0 0 0 OF TEXT ///more 4/>>Oe /4 reqrrr'red, Iree edd/rr'or>>l NRC Form 366AB/ (17)

Table 1 Parametric Evaluation of Secondary Containment/SGT Performance Sec. Roof Line Time To Outside Wind SGT Cont. Stdy State Reach Evaluation Description Temp Speed Flow Leakage Pressure -0.25"

('F) (mph) (cfm) (cfm) (AH20) (minutes)

Design Basis 'Performance of SGT and Secondary Containment 12 10. 3 4460 2240 -0. 02 Never Realistic Secondary Containment 12 1 0.3 4460 1475 -0.156 Never Leakage, Design Basis SGT Flow Design Basis Sec. Cont. Leakage, Realistic SGT Capability 12 1 0. 3 5600 2240 -0. 12 Never Realistic Sec. Containment and Realistic SGT Capability 12 10;3 5600 1475 -0.323 3.5 Reanalysis For Coldest Temperature Capability Realistic Sec. Containment and Realistic SGT Capability -23 5600 1475 -0.25 10 Realistic Sec. Containment and Realistic SGT Capability 10.3 5600 1475 -0.25 10 Reanalysis With 5X Margin Realistic Sec. Containment and Real istic SGT Capabil i ty 12 10. 3 5320 1475 -0. 282 Realistic Sec. Containment and Realistic SGT Capability 12 10.3 5600 1549, -0.295 3.6 On January 8, 1990, as a result of in-depth reviews of previous calculations pertinent to secondary containment accident performance, a problem with the meteorology data presented in FSAR Amendment 36 was discovered. It was determined that the values for the source term dispersion pattern relative concentration factor, X/g, were specified .incorrectly in FSAR Amendment 36. The error occurred as a result of incorrect input data used when the atmospheric dispersion calculation model was changed to comply with NRC Regulatory Guide 1.145.

NRC Form 366A (689)

NRC FORM 366A U.S. NUCLEAR AEGULATORY COMMISSION (6$ 9) APPROVED OMB NO. 31600104 E XP I R ES: 4/30/92 4ATED BURDEN PEA RESPONSE TO COMPLY WTH THIS LICENSEE EVE REPORT (LER) ES INFORMATION COLLECTION REQUEST: 500 HRS. FOAWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND AEPOATS MANAGEMENT BRANCH (P.530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1ME PAPERWORK REDUCTION PROJECT 131500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

LEA NUMBER (6)

Pip> SEOUENTIAL:~dC REVISION NUMBER NUMBER Washington Nuclear Plant TEXT /// mare epeoe /4>>r/u/red, u>> addio)roe/NRC %%dnn

- Unit 3//MS / ()7) 2 o 5 << o 3 9 7 9 0 0 4 0 0 1 OF A review, of the previosly completed JCO was conducted which resulted in the conclusion that, even with the correct X/g values inserted, both offsite and onsite post accident doses remain below 10CFR100 limits.'lthough this discovery does not present any new instance of reportability, this information is being provided on a voluntary basis as an update on the WNP2 Secondary Containment Performance problem originally reported in this LER. This information was also made available during a recent presentation made by Sypply System Generation Engineering to the Nuclear Regulatory Commission NRR Branch on January 16, 1990.

Immediate Corrective Action A Justification for Continued Operation (JCO) was performed for WNP-2. The conclusion of the JCO is that operation of the Plant can continue while final resolution of this issue is achieved.

On January 10, 1990, the previously prepared JCO was revised to include the effects of the corrected X/I) values. The conclusion of this revised JCO remains that the the operation of the Plant can continue while final resolution -is achieved.

I Further Eva1uati on and Correc ti ve Acti on A. Further Evaluation

1. This event is reportable under 10CFR50. 73(a)(2)(ii)(B) as a condition outside of the Plant design basis.

2. The cause of this event is design related in that inadequate design criteria were used by the Architect/Engineer (Burns and Roe, Inc. ) to determine SGT draw down time.

3. Current NRC requirements for radiological analysis do not allow SGT credit until a full -0.25H differential pressure is established at all Secondary Containment boundary surfaces. A review of existing radiological anlayses indicates that both the post-LOCA offsite and control room doses will increase as a result of delayed reestablishment (beyond two minutes) of the -0.25H differential. However, reanalysis using current rules (Standard Review Plan 6. 5. 5) that allow credit for iodine scrubbing within the suppression pool are expected to result in offsite doses equivalent to those outlined by the FSAR assuming a ten minute "no SGT credit" period to reestablish the full -0.25". The current condition of the SGT and Secondary Containment do not meet the FSAR description under all reasonable environmental conditions; however, the resultant doses are within the 10CFR100 and GDC 19 requirements.

NRC Form 366A (64)9)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6 J)9) APPROVED OMB NO. 31500104 EXPIRES: e/30/92 ES ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE REPORT (LER) INFORMATION COLLECTION REOUEST: 503) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P.530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31504104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

N SEOUENTIAL J~gP REVISION NUMSER vo< NUMSER Washin ton Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 9 0 0 0 OF TEXT /ifmore e/reoe le>>rloPed, rr>> eddlrlolre/NRC Form 3664'e/ (12)

4. Although there were no structures, components or systems inoperable prior to the event which contributed to the event, the equipment affected by this problem are SGT system filter trains SGT-FU-lA and SGT-FU-1B.

B. Further Corrective Action

l. This situation was reviewed relative to the requirements of 10CFR50. 59 and it was determined that it represents an unreviewed safety question.

Accordingly, the NRC was formally notified of this determination.

2. To confirm that the aforementioned actual Secondary Containment leakage value has remainea representative of the Plant condition, a test was run on September 26, 1989. The leakage was found to be 1228 cfm; thus, con-firming the 1475 cfm value used for the JCO.

3. Design Basis changes will be evaluated to provide an SGT system that allows for adequate filtering of Secondary Containment for applicable meteorological conditions, and system draw down time, so as to meet 10CFR100 and GDC 19 limits, while taking credit for suppression pool scrubbing as allowed by SRP 6. 5. 5.

4. Current system testing will be maintained to ensure Secondary Containment leakage and SGT flow capabilities are within the JCO analysis.

5. The NNP-2 FSAR will be revised to show the correct X/(} values.

Safet Si nificance Given the current state of Secondary Containment integrity, the SGT can provide adequate differential pressure control with an adequate margin applied for variations in Secondary Containment leak-tightness and SGT flow performance. Based upon actual Plant conditions arid system performance, the Secondary Containment pressure differential will remain greater than -0.25H during severely cold winter conditions; with temperatures as low as -23/F without wind and -8/F with a coincident 10 mph wind. Although formal calculations have not been prepared, preliminary calculations show that both offsite post accident doses remain well below 10CFR100 limits and, with credit for suppression pool scrubbing, not significantly different than the results now documented in the FSAR.

NRC Form 366A (6JIB)

NRC FORM 366A (64)9)

LICENSEE EVE TEXT CONTINUATION V.S. NUCLEAR REGULATORY COMMISSION REPORT ILER) t APPROVED OMB NO. 31600104 E XPIR E S: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 60A) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P.530), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (3(500104), OFFICE OF MANAGEMENTAND BUDGET,WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2I- LER NUMBER (6) PAGE (3)

YEAR gj% SEQVE NTIAL >~~ REVISION NUMSER "<YA NUMBER Washin ton Nuclear Plant - Unit 2 o s o o o 39 90 0 0 01 8 o"0 8 TEXT /llmore SOece /4 ler)o/red. fee edde'one/ NRC Forrrr 3664'4/ ()7)

In light of the X/Q input errors discovered on January 8, 1990, the Unreviewed Safety Question Analysis originally prepared was reviewed and revised. Although the errors in FSAR methodology compound the offsite dose consequences, the study calculations performed to assess the impact of this discovery continue to support the conclusions of the original analysis. The conclusion remains that both the offsite and onsite post accident doses are within the 10CFR100 limits using the correct X/Q values in the atmospheric dispersion model.

Similar Events LER 88-023 EIIS Information Text Reference EIIS Reference System Component Standby Gas Treatment (SGT) System BM Secondary Containment NG Standby Service Water (SSW) BS Emergency Diesel Generator EK GEM Reactor Building Exhaust and Outside Air (REA and ROA) Isolation Valves VA ISV SGT-FU-lA and SGT-FU-1B BM FLT NRC Form 366A (64)9)