ML17286B100

LER 91-024-00:on 910909,unanalyzed Condition Associated W/ Postulated Main Steam Line Failure Outside Containment Discovered by Ge.Caused by Failure to Consider Iodine Source Term.Procedure Re Cold Startup changed.W/911010 Ltr

ML17286B100
Person / Time
Site:	Columbia
Issue date:	10/10/1991
From:	BAKER J W, FIES C L WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GO2-91-183, LER-91-024, LER-91-24, NUDOCS 9110240006
Download: ML17286B100 (11)

Text

T D DISTRIBUTION DEMONSTRATION SYSTEM~~REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9110240006 DOC.DATE: 91/10/10 NOTARIZED:

NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH.NAME AUTHOR AFFILIATION FIES,C.L..Washington Public Power Supply System BAKER,J.W.

Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 91-024-00:on 910909,unanalyzed condition associated w/postulated main steam line failure outside containment discovered by GE.Caused by failure to consider iodine source term.Procedure re cold startup changed.W/911010 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED'LTR L ENCL 3 SIZE TITLE: 50..73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES: RECIPIENT COPIES'ECIPIENT ID CODE/NAME LTTR ENCL ID CODE/NAME PD5 LA 1 1 PD5 PD ENG,P.L.1 1 COPIES LTTR ENCL 1 1 D INTERNAL: ACNW AEOD/DOA AEOD/ROAB/DS P NRR/DET/EMEB 7E NRR/DLPQ/LPEB10 NRR/DREP/PRPB11 NRR/DST/SICB8H3 NRR/DST/SRXB 8E RES/DSIR/EIB EXTERNAL: EG&G BRYCE,J.HNRC PDR NSIC POORE,W.2 2 1 1 2 2 1 1 1 1 2 2 1 1 1 1 1~1 3 3 1 1 1 1 ACRS AEOD/DS P/TPAB NRR/DET/ECMB 9H NRR/DLPQ/LHFB10 NRR/DOEA/OEAB NRR/DST/SELB 8D ST/PLB8D1 REG ILE 02 FILE 01 L ST LOBBY WARD NSIC MURPHY,G.A NUDOCS FULL TXT 2 2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 D NOTE TO ALL"RIDS" RECIPIENTS:

D D PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOii I P 1-37 (EXT.20079)TO ELIMINATE YOUR NAME FROiVI DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 33 ENCL 33 WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O.Box 965~3000 George IVashinglon 1Vay~Richland, 1Vashington 99352 Docket No.50-397 October 9, 1991 602-91-183 Document Control Desk U.S, Nuclear Regulatory Commission Washington, D.C.20555

Subject:

NUCLEAR PLANT NO.2'ICENSEE EVENT REPORT NO.'91-024

Dear Sir:

Transmitted herewith is Licensee Event Report No.91-024 for the WNP-2 Plant.This report is submitted in response to the report requirements of 10CFR50.73 and discusses the items of repor'tability, corrective action taken, and action taken to preclude recurrence.

Very truly yours, t>J.W.Baker WNP-2 Plant Manager

Enclosure:

Licensee Event Report No.91-024 cc: Hr.John B.Hartin, NRC-Region V Mr.C.Sorensen, NRC Resident Inspector (H/D 901A)INPO Records Center-Atlanta, GA Ms..Dottie Sherman, ANI Hr.D.L.Williams, BPA (M/D 399)NRC Resident Inspector-walk over copy y 4$0 ay qO()/, 9$j (.1'I 0 pp,"-!QOPji V, A5000-97 F'DR+pzz NRC FOAM 366 (64)9)V.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT ILER)APPROVE O OMB NO.31500104 EXPIRES: 4130192 EO>IMATEO BURDEN PER RESPONSE TO COMPLY WTH THIS INFOAMATION COLLECTION REOUEST: 50.0 HAS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (PJ)30).U.S.NUCLEAR REGULATOAY COMMISSION, WASHINGTON.

OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104).

OFFICE OF MANAGEMENT ANO BUDGET, WASHINGTON, DC 20503.FACILITY NAME (11 Washington Nuclear Plant-Unit 2 TITLE (4)DOCKET NUMBER (2)PAGE 3 0 5 0 0 0 3 9 7 1 OF 0 Unanalyzed Condition Associated With Hain Steam Line Failure Outside Containment EVENT OAT E (6)LER NUMBER (6)REPORT DATE (7)OTHER FACILITIES INVOLVED (8)MONTH DAY YEAR YEAR'2 SEQUENTIAL

.9 NUMBER REVISION'OW NUMBER MONTH DAY YEAR FACILITY NAMES DOCKET NUMBFR(S)0 5 0 0 0 0 9 0 9 9 1 9 1 024 0 0 10 0991 0 5 0 0 0 OPERATING MODE (9)POWER LEYEL 0 0 0 (10)20.402(B)20.405(eHIHB 20.405 (e I (1)(ii)20.405 (~l(1)(ill)20.40d(~HI Hlvl 20.405(eHIHv) 20.406(cl 50.3d(cHI) 50.3d(c)(2)d0.73(e)(2)I I)50.73(e H2)(ill 50.73(eH2Hiiil LICENSEE CONTACT FOR THIS LER (121 60.73(e)(2)Hv)60.73(el(2)(v) 50.73(e)(2)(vil)50,73(~H2HvlilHA) 60.73(~l(2)(vill)(B) 60,73(~)(2)lel THIS REPORT IS SUBMITTED PURSUANT T 0 THE RtOVIREMENTS OF 10 CFA (): (Check one or more of the foifowinpl (ill 73 71(B)73.71(cl OTHER ISpecily in Ahrtrert hefovrend In TecL NRC Form 366AI NAME C.L.Fies, Compliance Engineer TELEPHONE NUMBER AREA CODE 50 937 7-4147 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT ILS)CAUSE SYSTEM COMPONENT MANUFAC TVRER EPORTABLE.e:C 44~+<gM TO NPADS'Q~p..ly((y., CAUSE SYSTEM COMPONENT MANUFAC.TUAER EPORTABLE TO NPRDS k4@~%'UPPLEMENTAL REPORT EXPECTED (14)YES Ill yN, compiere EXPECTED StlperiSSION DATEI NO ABsTRAGT I(.lmit to fepp rpecer.i.e., epproeimerery fifteen rinpie Ipece rypervrrtren lined (16)EXPECTED SUBMISSION DATE IISI mr)i@gk(vir MONTH DAY YEAR ABSTRACT On September 9, 1991 a reportabi lity evaluation was approved which concluded that an unanalyzed condition associated with a postulated main steam line failure outside containment had existed at WNP-2 during prior operating cycles.This condition was discovered by General Electric while performing a recalculation of the accident as a result of updated meteorological data.The unanalyzed condition was caused by the need to'onsider an additional iodine source term as a result of the postulated mass release through main steam line drains in the event of a high energy pipe break outside the primary containmeni..

This unique accident scenario results in Reactor Pressure Yessel (RPY)depressurization and an increased radiological source term due to iodine spiking that is not usually associated with a main steam line break.NRC Form 366 (54)9)

NRC FORM 356A (669)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVE EPORT (LER)TEXT CONTINUATION APPROYEO 0MB NO, 31504104 EXPIRES: 4/30/92'ST.ED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 500 HRS.FORWARD'OMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P 530), U S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3)504)104).

OFFICE OF MANAGEMENT AND BUDGE1'.WASHINGTON, DC 20503.FACILITY NAME (1)DOCKET NUMBER l2)LER NUMBER (6)YEAR<<Pi SEOVENTIAL PA REVSION NUMBER..P5 HUMBE+PAGE (3)Washington Nuclear Plant-Unit 2 osooo1 024 0 0 0 2oF 0 9 TEXT/I/mom pwca Ss m9oirrd, wt addi(iona/

NRC Form 366AB/l 12)Since the plant was is mode four (cold shutdown)no immediate corrective action was required for the unit itself.Immediate corrective action was taken to change Plant Procedure PPH 3.1.2, Reactor Plant Cold Startup', to require the five drain valves located outside containment (MS-V-67A, B, C, D, and HS-V-19)to be closed whenever reactor power is greater than or equal to five percent.The root cause of this event was less than adequate change management.

The risks and consequences associated with the change in operating procedures was not adequately reviewed and assessed.Further corrective actions will be taken to assure personnel are aware of the need to carefully control changes.Calculation procedures will be strengthened and a review will be performed to assure adequate controls are in place for calculations done by internal organizations.

The safety significance review showed the impact of the postulated accident, had it occurred with depressurization, would have been greatly reduced because of the very small number of failed fuel rods during past operating cycles.In addition, downstream valves, even though not safety related, would have most likely been available to stop the long term depressurization.

The event posed no threat to the health and safety of either the public or plant personnel.

Plant Conditions Power Level-0 A Plant Mode-4 Event Descri tion On September 9, 1991 a reportabi lity evaluation was approved which concluded that an unanalyzed condition associated with a postulated main steam line break (MSLB)outside containment had existed at WNP-2 during prior operating cycles.This condition was discovered by General Electric while performing a recalculation of.the FSAR Chapter 15 accidents and transients as a result of updated meteorological data.The unanalyzed.condition involved the amount of water and steam released in the event of a MSLB outside the primary containment.

The current accident analysis, as descr.ibed in Section 15.6.4 of the FSAR, assumes a break in a main steam line downstream of an outboard isolation valve.A single failure.of one of the inboard Hain Steam Isolation Valves (HSIV)is also assumed in the analysis.Within seconds, however,*the remaining outboard MSIV closes, and the release.is limited to the mass flowing through the valve while it is closing (119,000 pounds).This current FSAR accident scenario does not result in significant primary system depressurization.

NRC Form 366A (669)

NRC FORM 366A (64)9)US.NUCLEAR REGULATORY COMMISSION LICENSEE EVEN EPORT ILER)TEXT CONTINUATION P APPROVEO 0MB NO.3I504))04 E X PI RES)4/30192 ESTI O BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 500 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P4)30).U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON.

DC 20555.AND TO THE PAPERWORK REDUCTION PROJECT 131500104).

OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON.DC 20503.FACILITY NAME (I)DOCKET NUMBER (1)YEAR LER NUMBER (6)SEQUENTIAL NUMBER'~r(S rsvMosA PAGE (3)Washington Nuclear Plant-Unit 2 TEXT Ul more spsse is rsr(rrr)ed, ose eddroorrsl A'RC Fonrr 3654's)(IT)o s o o o 3 9 024 0 0 0 3 op 0 9 The condition discovered involves additional mass release through main steam drain line piping under the above accident scenario.These drain lines are isolated outside primary containment by five main steam 1 ine drain valves (HS V 67A, 8, C, D, and NS-V-1 9).During heatup and cool down condi t ions the drai'n lines that each contain a 67 valve provide a means of removing moisture from their associated main steam lines'in the area between the inboard and outboard isolation NSIVs.The fifth drain line that contains HS-V-19 is designed to remove moisture from all main steam lines upstream of the'inboard isolation valves.During the postulated accident conditions jet impingement from the steam line break outside containment could cause these valves to fai 1"as-is" at the time the accident occurred.During the past operating cycles (since early 1984)WNP-2 has been operating with these valves.open.If the main steam line break accident had occurred, as described in the FSAR, additional mass could be released through these valves.The highest mass release would occur if blowdown occurred through the line containing NS-V-19 since this is a three inch line while the lines containing the 67 valves are one and one half inch diameter lines.A single failure, plus the consequences of the jet impingement, even on all the HS-V-67 valves and: the HS-V-19 valve, can only result in one unisolaied blowdown path being available.'he single failure is either the failure of an inboard NSIV to close (leading to blowdown through one MS-V-67 valve)or the fai lure of NS-V-16 (leading to blowdown through NS-V-19).Any of these lir)es could result in long term depressurization of the primary system if isolation by downstream non-safety related valves could not be achieved.The depressurization could also result in iodine release from the failed fuel assumed to be present just prior to the event.This iodine release would cause"iodine spiking" usually associated with instrument.

line breaks in boiling water reactors and steam generator tube ruptures in pressurized water reactors.The long term depressurization aspect of the event described above makes it different from the accident desscribed in Standard Review Plan 15;6.4, Radiological Consequences of Main Steam Line Failure Outside Containment (BWR).Consequently, this scenario would increase the radiological source term for the calculation beyond that evaluated in the FSAR resulting in an unanalyzed condition.

Immediate Corrective Action Since the plant was in mode four (cold shutdown)no immediate corrective action was required for the unit.Immediate corrective action was taken to change Plant Procedure PPN 3.1.2, Reactor Plant Cold Startup, to require the drain valves to be closed whenever reactor power is greater than or equal to 5 percent.NRC Form 366A (64)9)

NRC FORM 366A (6419)U.S.NUCLEAR AEGULATORY COMMISSION LICENSEE EVE REPORT ILER)TEXT CONTINUATION

.APPROVED OMB NO.3(504)(04 EXPIRES: 4I30/92 ES EO BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND AEPORTS MANAGEMENT BAANCH (F430), U.S.NIJCLEAR REGULATOAY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK AEDUCTION PROJECT (31504)I04), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.FACILITY NAME II)DOCKET NUMBEA (2)(.ER NUMBER (6)PAGE (3)YEAR pe SEDVENTIAL

'dip NVMBER REVIS10N NUMSER Washington Nuclear Plant-Unit 2 TEXT III more spree (4 reohhired, o>>edChrhorhei NRC Form 3MA's)(17)o so oo397.1 024 0 0 0 4 OF 0 9 Further Evaluation and Corrective Action A.Further Evaluation l.This event is being r'eported per the-requirements of 10CFR50.73(a)(2)(ii)(B) as".....a condition that was outside the design basis of the plant.~...".2.Further evaluation of this event showed a long history associated with this issue: a~The original design of the plant and the supporting analysis done by General Electric prior to 1983 assumed the drain valves were closed (or could be closed)under accident condi Lions.The Plant Operating Procedure in effect in mid 1983, PPH 3.1.2, Reactor Plant Cold Startup, Revision 4, implemented this design requirement by requiring the valves to be closed when the plant was above 5X power.Application of the rules for high energy line break analysis allow a va')ve to be considered"closed" if it is closed at or below 5X rated power.b.In Oecernber 1983 Burns and Roe, the Architect-Engineer for WNP-2, complei,ed the stress analysis on the drains from the main steam lines and determined that they could not be certified for the life of the plant if the valves were closed at greater than 5$power and reopened at below 5X power.They recommended adding orifices to the lines and letting them blowdown continuously in order to avoid the severe thermal stress cycles.This recommendation was implemented by a change to PPH 3.1.2 (Revision 5)in April 1984.It did not appear that the personnel making the change to the procedure w'ere aware of the requi rements of the design and the accident analysis and the plant was operated with the valves open until the refueling outage in 1985.C.In Hay 1985 a nonconformance report (285-0234) was generated when it was determined that both FSAR, section 3.6, and Burns and Roe calculat,ion 5.51.059 requ.ired the drain valves to be, closed at greater than 5%%d power.The nonconformance

'eport again raised a concern about the thermal stress on U)e valves.As a result of this concern, a evaluation was performed to evaluate the radiological NRC Ferro 366A (689)

NRC FOAM 366A (649)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVE REPORT (LER)TEXT CONTINUATION APPROVEO 0MB NO.3'I50010l EXPIRES: l/30(92 ES ED BURDEN PER AESPONSE TO COMPLY>WTH THIS INF ATION COLLECTION AEOUESTI 50Aj HRS.FOAWARD COMMENTS REGARDING BUAOEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BAANCH (P-530), U.S.NUCLEAR AEGULATOAY COMMISSION, WASHINGTON.

DC 20555, AND TO THE PAPERWOAK REDUCTION PROJECT (31500'Iol).

OFF ICE OF MANAGEMENT AND BUDGET, WASHINGTON.

DC 20503.FACILITY NAME (1)DOCKET NUMBER (2)YEAR LER NUMBER (5)SEOUENTIAL N U M 1 E R Jeer REVISION NUMBER PAGE (3)TEXT (if mors spsco is rsqrdrod, oso odds'onsiiVRC fonrr 3554's)(17)0 5 0 0 0 OF J consequences of operating with the valves open.This evaluation was perFormed by one of the plant support organizations, Radiological Programs, at the request of Engineering.

The evaluation assumed an additional 1,080,000 pounds oF saturated liquid and steam was released because of the open drain valves but the radiological source term was taken from the FSAR analysis.With the assumptions used, the calculated doses were all less than one rem and, based on these results, the decision was made to operate above 5X power with the drain valves open.The plant continued to operate in this condition until the refueling outage in 1991.d.In early 1991 General Electric was contracted to recalculate the off-site consequences of the main steam line failure outside containment accident.This was done to incorporate revised meteorology that was added to the FSAR in the 1990 amendment.

When it was learned by General Electric that the reason for the high steam and liquid mass losses prov.ided by the Supply System for this event was that WNP-2 was operating with the drain valves opened it was pointed out that the increased source term due to iodine spiking needed Ko be included in the evaluation.

A preliminary calculation was performed by General Electric using a non-production program to estimate the dose.In this evaluation the total iodine spiking source term calculated from the instrument.

line break accident (WNP-2 FSAR Section 15.6)was assumed to be released to the reactor pressure vessel immediately following a main steam isolation valve closure.This is a conservative assumption since the normal spiking source term released to the RPV is proportional to the rate of vessel depressurization.

Also, as the RPV depressurizes less of the mass loss will be saturated steam and, ti)erefore, less iodine will be released.Further, since the mass of coolant lost through the drain lines is in excess of the coolant initially in the RPV it is assumed tt)at all the spik'ing activity is released to the environment.

With these assumptions the calculated maximum off:site thyroid dose is 239 rem which is within the 300 rem limit of 10CFR100.The evaluation by General Electric was not finalized with a production program since the decision was made to operate with the valves closed.3.Further investigatior) found that ti)e evaluation done in 1985 did not use the formal process established by Radiological Programs for performing calculations to be used for safety related applications.

The memo transmitting the results of the evaluation recommended a refined analysis be performed using codes referenced in the FSAR.Radiological Programs Instruction (RPI)=Manual Procedure RPI 2.1, Evaluation Log, should have been used to perform the evaluation for this application.

NRC Form 366A (669)

NRC FORM 366A 1609)US.NUCLEAR REGULATORY COMMISSION LICENSEE EVE EPORT (LER)TEXT CONTINUATION APPROVED OMB NO.31500)Of EXPIRES: 4/30/92 D BURDEN PER RESPONSE TO COMPLY WTH THIS INFO ATION COLLECTION REQUEST: 50JI HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH IP430), U.S.NUCLFAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT 13)500104).

OFFICE OF MANAGEMENT AND BUDGET.WASHINGTON, DC 20503.FACILITY NAME Ill DOCKET NUMBER (2)LER NUMBER (6)PAGE 13)Washington Nuclear Plant-Unit 2 TEXT/I/mme Jtwco/f nqotrod, oso oddrtioIM!

HRC FooII 36649/Ill)o s o o o3 YEAR Af VISION IIUM9 f 4 024 0 0 0 6 SEQUENTIAL 954 NUMINE 4;Yiv'.OF 4.Further evaluation did not find any record of a Generation Engineering review of the results of the evaluation.

Generation Engineering had formally requested the evaluation be performed by memo and had provided the mass release value.However, since the formal calculation was not completed and transmitted to Engineering no review was performed.

The only record of review was the nonconformance report which used the Radiological Programs memo as justification for closeout of the NCR.There was r)o Engineering signoff of the NCR as it was not required by the process at that time.Generation Engineering believed a formal calculation would be performed by Radiological Programs and delayed changing the FSAR to reflect the open valves until it was received.However, Radiological Programs was not aware of the need for a formal calculation.

Tracking systems in place at that time were not sufficient to flag this item for management attention.

5.A 50.59 review of the change to Plant Procedure PPN 3.1.2.Reactor Plant Cold Startup, was performed in March 1984.The 50.59 review should have recognized the significance of the change which allowed operation with the drain valves open, 6~The root.cause of this event was less than adequate change management.

The risks and consequences associated with the change in operating procedures was not adequately reviewed and assessed.The change management process should have detected this deficiency in 1984 when the initial change to the procedure was made and again in 1985 when the non-conformance report was signed off with a use-as-is disposition.

This event also had three contributing causes.Required procedures were not used by Radiological Programs in performing the radiological analysis for this safety related application.

In addition, there was inadequate communication between organizations on the end use and si:atus of the evaluation.

Finally, personnel performance was a contributing cause.The formal calculation should have been done in a timely manner and, failing that, the issue should have been brought to management attentions There were no structures, components or systems that were inoperable prior to the start of this event which contributed to the event.Further Corrective Action A lessons learned evaluation will be generated as part of the root cause analysis and this will be made a part of the training or required reading for all Plant 1'echnical, Plant HP/Chemistry, Generation Engineering, and Radiological Programs personnel performing safety related calculations.

NRC FoII4366A 1669)

NRC FORM366A (64)9)l US.NUCLEAR REGULATORY COMMISSION LICENSEE EVEN EPOR7 ILER)TEXT CONTINUATION APPROVEO OMB NO.31504(04 EXPIRES: 4I30)93 ESTI D BUADEN PER RESPONSE TO COMPLY WTH THIS INFOAMATION COLLECTION AEOUESTI 50JI HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND AEPORTS MANAGEMENT BRANCH (F430), U.S.NUCLEAR AEGULATORY COMMISSION.

WASHINGTON, DC 30555, AND TO THE PAPERWORK AEOUCTION PROJECT (3'l50d)04).

OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, OC 30503.FACILITY NAME ill DOCKET NUMBER (2)YEAR LER NUMBER (6)SSOUENTIAL NUMBER 2.W REVISION A:.5'5 NUMBER PAGE Ll)Washington Nuclear Plant-Unit 2 TEXT Iil IIOIO SOsCO Js sooowsd, vso oddio'oooJ HRC Form 3664'si ()T)0 5 0 0 0 3 9 7 024-0 0 0 7"0 9 2.Radiological Health Instruction 2.1, Calculation Logs, will be amended to ensure the intended use of the evaluation is clear to the individual performing the evaluation and that the use is identified within the body of the evaluation.

3.An adequacy review will be performed of the present policies and processes used for calculations by internal organizations supporting WNP-2 Oesign Engineering.

4.The'impact of thermal cycling on the drain valves and lines will be reviewed to assure design limits wi 11 not be exceeded.5.No further corrective action is believed to be necessary for the 50.59 and change management process, 1'he process has been strengthened significantly since this event occurred in the 1984-1985 time frame.Methods of identifing and tracking corrective actions have been improved.This includes additional training that is ongoing at this i.ime.This should prevent reoccurrence of this problem in the fui.ure.6.No furi,her corrective action is needed for tracking outstanding items as this process has also been s.Lrengthened to assure periodic reviews are performed by management in a timely manner.Safet Si nificance The plant has been operating outside the bounds of the analysis as described in Chapter 15 of the FSAR.However, one factor that would mitigate the significance of this condition involves the capability for isolation using'the downstream valves{see attached figure).The long term depressurization resulting from the NS-V-19 or-a NS-V-67 valve failing open during a NSLB could have been prevented by closure of NS-V-21 and NS-V-156{for NS-V-19)or AS-V-68 and NS-V-69{for a NS-V-67 valve).Credit is not taken for these valves and associated piping in the acc'ident analysis as they are not safety related.An additional important factor that would decrease the significance of the NSLB, if it had occurred, would be the condition of the fuel at WNP-2 during past operating cycles.The number of fuel rod failures has been very small.The iodine spiking occurs when the gas trapped within a failed rod during power operation is released on depressurization

~The accident analysis recently performed by General Electric that resulted in the calculated thyroid dose of 239 rem assumes a fuel failure fraction at the start of the accident that was typical of BWR reactor experience more than 10 years ago.At WNP-2 there have been only six rod failures since plant s tartup.The worst case fuel cycle involved four pin failures.Therefore, if a NSI.B accident had occurred there would have been signif'icant releases of primary coolant but the wad.iologica1 consequences would have been below 10CFR100 1 imi t.s because of the low amount of fuel failures.NRC FonII 366A (669)

NAC FOAM06&A (689)U.S, NUCLEAA AEGULATORY COMMISSION LlCENSEE EVEN REPORT (LER)TEXT CONTINUATlON APPAOVEO 0MB NO.3)500106 EXPIRES: S/30/92 EST, EO BURDEN PER RES)'ONSE TO COMPLY WTH THIS INFOAMATION COLLECTION REQUEST: 50/)HAS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS ANO AEPOATS MANAGEMENT BRANCH (F430), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWOAK REDUCTION PROJECT (31504104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.FACILITY NAME (II DOCKET NUMBER (2)YEAR LER NUMBER (6)SEQUENTIAL MC;NUMSER REVISION NUMSEA PAGE (3)Washington Nuclear Plant-Unit 2 TEXT/i/move Aoecm/r rN/rn'red, rrro eddroSrrel llRC Fcvm 3/JIS42/(12) o s o o o 024 0 0 0 8 oFo Similar Events There are no similar events.EIIS InFormation Text Reference EIIS Reference~Sstem~Com anent Main Steam System SB Main Steam Drain Valves (HS-V-67A, B, C, D, and NS-V-19)SB NRC Form 36&A (669)

?A 0'll A I Z ez*D mn 0 D 3 dl nEAcron (R(VACUUM SAEANEns~ACIC 7+tl I tti tt+0 IIV ln 414 AM snv TAILPIPE IIVP OF Id)MD 14 AEACTOA BUILOIHO 14 14 td MD MD MO MO 41 41 41 20 V%0 MO 44 MD 21 MD 10 V.I$4 TO TUADDIE Collrno SYSTEM 0 420+~VI pr~F11 I C~IC To MSLC MO.V ti 144 MO V.22 V-12 FAOM ACIC STEAM LINE~Dnun IIO 140 YO MAIM COHDEtlSEll IIO I 40 MD 10A MD 10 MO 100 To SEAL STEAM EVAPOAATORS To AF P TUIIOINES To MSird tmt STAO'E IIENEATEAS MO IOD TOH TunoiNE TUnoiNE Tnnott t~l OOVBUIOII VALVE ASSEMOLIES TOHPTUAOINE TOOFFNAS PAEHEATEAS To SJAE'4 I.O rt ID OI O'f c+m Z C m D A+m lTT+x~~ITT A~9 O~-0 2 c fTT n Xl m D m 0 C 0 D'C A 0 ron DOWIICOMEA OAYWELL SUPPAESSION VOLUME NOTEi ALL VALVES AnE MS VALVES UNLESS OIIIEAWISE NOIED.FIGURE 1.MAIN STEAM SYSTEM IO I I IO 140 YO MAIM COtlDEIISEA

'C D OI ZO CC g III mz l r I@0 n C III 0(mn III Q DZ I Z C m D 0 m 0 m c a m O D 0 Ill~n nl A ln 0 g 6 0 0>DNA 3moo3o~~CD3D 3+0m~m mD D 0 3+D~Dom m O+ln m Z mC Zm3P Z 0 00m ln O?O?0 D mclnmS~m CA03moD oo OH?C ZD?>me?D 0 Z Z vl min ln\Zmn-nm?3 0 ZDqvlo zmo D" HQ Simon 0 op~120 Z+A.0?3~O~A-5.Ztn 06'9CWDg Qv mm'in?g'Dcmg ln 0 Z A n g Z'0mDDH n<+ODZ moDcnOQ 0?