IR 05000266/1992010
| ML20046D467 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/04/1993 |
| From: | Hansen A Office of Nuclear Reactor Regulation |
| To: | Link R WISCONSIN ELECTRIC POWER CO. |
| References | |
| RTR-NUREG-CR-4674 NUDOCS 9308200162 | |
| Download: ML20046D467 (8) | |
Text
{{#Wiki_filter:! v i August 4, 1993 j . Docket Hos.
50-266 DISTRIBUTION: I and 50-301 Docket File OGC NRC & Local PDRs ACRS (10) Mr. Robert E. Link, Vice President PDIII-3 Reading KJury, SRI Nuclear Power Department JRoe FManning, AEOD Wisconsin Electric Power Company JZwolinski Region III, DRP 231 West Michigan Street, Room P379 JHannon PDIII-3 Gray File Milwaukee, Wisconsin 53201 MRushbrook TMcGinty AHansen
Dear Mr. Link:
GHolohan, AEOD SUBJECT: REQUEST FOR COMMENTS ON THE PRELIMINARY DRAFT OF NUREG/CR-4674, " PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS: 1992, A STATUS REPORT" By letter dated June 29, 1993, you provided comments on a voluntary basis on the subject draft NUREG Section B.14, based on Point Beach Nuclear Plant (PBNP), Unit 2 Licensee Event Report (LER) 92-003.
We appreciate the timely submittal of your comments and they are currently under review. Additionally, three other operating events were still being analyzed as possible precursors at the time of our first request, one of which involved PBNP LER 266/92-010.
Enclosed is a copy of an accident sequence precursor (ASP) analysis of Point Beach Licensee Event Report (LER) No. 266/92-010 performed by the Oak Ridge
National Laboratory (ORNL) contracted by the NRC. The ASP analysis will be ' part of NUREG/CR-4674 scheduled to be published in September 1993.
We are interested in any comments you may have on this preliminary analysis.
Your response, which is strictly voluntary, should address the following three areas; report characterization of conceivable plant responses to the event, representation of safety system configuration, and analysis assumptions
regarding equipment recovery. Any responses received by August 31, 1993, will -; be consit. ed by AE0D and ORNL in the final draft of NUREG/CR-4674.
We recognize thai +he timeliness of this response may be an inconvenience, and genuinely appreciate your efforts to work with us.
Should you have any
questions, please contact me at (301) 504-1390.
r !
Sincerely, Robert Stransky/for Allen G. Hansen, Project Manager Project Directorate III-3
Division of Reactor Projects III/IV/V
Office of Nuclear Reactor Regulation ' Enclosure: As stated cc: See next page O G 00T.
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d UNITED STATES ., o . [j"; kZf j i NUCLEAR REGULATORY COMMISSION ( j WASHINGTON, D.C. 20555-0001 i \\. August 4, 1993 ' . . Docket Nos.
50-266 and 50-301 i Mr. Robert E. Link, Vice President ' Nuclear Power Department , Wisconsin Electric Power Company ' 231 West Michigan Street, Room P379 , Milwaukee, Wisconsin 53201
Dear Mr. Link:
SUBJECT: REQUEST FOR COMMENTS ON THE PRELIMINARY DRAFT OF NUREG/CR-4674, " PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDEhTS: 1992, A ! STATUS REPORT" By letter dated June 29, 1993, you provided comments on a voluntary basis on the subject draft NUREG Section B.14, based on Point Beach Nuclear Plant (PBNP), Unit 2 Licensee Event Report (LER) 92-003. We appreciate the timely submittal of your comments and they are currently under review. Additionally, three other operating events were still being analyzed as possible precursors at the time of our first request, one of which involved PBNP LER 266/92-010.
Enclos H is a copy of an accident sequence precursor (ASP) analysis of Point Beach Licensee Event Report (LER) No. 266/92-010 performed by the Oak Ridge National Laboratory (ORNL) contracted by the NRC. The ASP analysis will be part of NUREG/CR-4674 scheduled to be published in September 1993. We are .' interested in any comments you may have on this preliminary analysis.
Your response, which is strictly voluntary, should address the following three " areas; report characterization of conceivable plant responses to the event, ! representation of safety system configuration, and analysis assumptions regarding equipment recovery. Any responses received by August 31, 1993, will be considered by AE0D and ORNL in the final draft of NUREG/CR-4674. We recognize that the timeliness of this response may be an inconvenience, and genuinely appreciate your efforts to work with us.
Should you have any questions, please contact me at (301) 504-1390.
Since ,, su e= . ra > " Illen G. Hansen, Project Manager
Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear keactor Regulation ' Enclosure: , As stated
CC: I See next page l
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Mr. Robert Point Beach Nuclear Plant . Wisconsin Electric Power Company Unit Nos. I and 2 r CC: ! Ernest L. Blake, Jr.
i Shaw, Pittman, Potts & Trowbridge f 2300 N Street, N.W.
{ Washington, DC 20037 ' , Mr. Gregory J. Maxfield, Manager f Point Beach Nuclear Plant Wisconsin Electric Power Company 6610 Huclear Road ' Two Rivers, Wisconsin 54241 Town Chairman Town of Two Creeks
Route 3 Two Rivers, Wisconsin 54241 ' Chairman
Public Service Commission of Wisconsin . Hills Farms State Office Building
Madison, Wisconsin 53702 ' , Regional Administrator, Regicn III , U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn,' Illinois 60137 Resident Inspector's Office U.S. Nuclear Regulatory Commission . 6612 Nuclear Road . Two Rivers, Wisconsin 54241 't ! . . i
. . . .
.- . . - Mostire ' PRELBIINARY ' . A.1 LER Number 266/92-010 Event Description: Stiety injection system unavailable during testing Date of Event: December 8,1992 Plant: Point Beach I and 2 A.1.1 Summary Two quanerly valve stroke tests were found to isolate the minimum flow recirculation line common to both safety injection (SI) pumps. If the pumps were demanded when the recirculatics line was isolated, pump failure would quickly occur for high reactor coolant system (RCS) pressure conditions. De conditional core damage probability estimated for this event is B.3 x 104 His probability la applicable to both units. De relative significance of this event compared to other postulated events at P7 int Beach is shown in Fig. B.I.
IER 26692410 1E 7 IE-6 1E-5 1E4 1E-3 1E-2 I I J l I I i ^ i!i precurmr cutoff -~3 kNg-1.OOP - - 360h FP 360um Fig. B.I. Relative event significance of LER 266/92-010 compared with other potential events at Point Beach.
A.I.2 Event Description Point Beach Units I and 2 were at 100% and 95% power, respectively, on Dm; ember 8,1992. De utility discovered that inservice Tests IT 40, " Safety injection Valves (Quarterly), Unit 1," and IT-45, " Safety injectior. Valves (Quarterly), Unit 2,* isolate the SI pump common ninimum flow recirculation line by closing valves 397A and 897B. He utility stated that operation of the SI pumps at high RCS pressure conditions with the minimum recirculation line isolated would result in pump failure in ~ l min.
LER NO: 266/92-010 B-1 PRELBIINARY .- - - .- - - - - -
. PRELIMINARY ' A review of station logs indicated that the time required to mmplete ITd0 and TT-45 is on the order of two hours; however, the recirculation line is not isolated for Qe full duration of the test.
A.I.3 Additional Event-Related Inforrnation ) , An orificed minimum flow bypass line is p ovided at the discharge of each SI pump to recirculate tiow to the refueling water storage tank (RWST) through a common header, or mmtmum flow recirculation line, in the event that the SI pumps are run while the RCS preaure is above the pumps' shutoff head.
These bypass lines also permit the performance of periodic surveillance tests required by the Te&nical
Specifications to demonstrate pump operability. The recirenlation line is provided with series air-operated isolation valves 897A and 897B, which are c!csed to prevent the transfer of containment sump inventory i to the RWST during the rec:rculation phase following a lossef-coolant accident (LOCA). Because valves 897A and 897B fail close i, they are normally gagged open to prevent closure on a loss of instrument air.
' If the Si pumps are opera.ed without a flow path, the pumps will overheat and quickly deteriorate.
Valves 897A and 897B ar t interlocked with the containment sump isolation valves. These valves are normally closed except duting the recirculation phase following a LOCA. The interlock ensures the sump isolation valves cannot De opened until valve 897A or 897B is closed.
, A.I.4 Modeling Assurnptions The event has been modeled as an unavailability of high-pressure injection (HPI) and feed and bleed for an 8-h period within a 1-yr observation period (the interval between precursor reports). All small-break LOCAs were assumed to slowly depressurize, such that RCS pressure would remain above the shutoff head of the SI pumps long enough to fail the pumps. Because of the short time period before Si pump failure, no recovery was assumed possible.
A.1.5 Analysis Results The conditional core damage probability estimated for the event is 8.3 x 10*. The dominant core damage sequence, highlighted on the event tree in Fig. B.2, involves a potential small-break LOCA with failure of HPl. As noted above, the minimum flo v recirculation valves are not closed for the entire 2-h quarterly test period, and therefore this analysis is haewhat conservative.
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l ' LER NO: 266/92-010 . B-2 PRELBfINARY .
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i . f LER NO: 266/92-010 , B-3 PRELIMINARY i !
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L CDICITIONAL CORE DAMAGE PROBASILITT CALCULATI0ts I Event IdontifIer 266/92-010 Event Descriptient safety injection system unevellebte skring testing Event setes - 12/08/92 ! Plants Point Beech 1 ' 13NAVAILA81LITY, DURATIoss 8 NON RECWERABLE INITIATANG EVElff PROEARILITIES , TRANS 5.1E 04 LOOP 4.7E-05.
LOCA 8.3E-06 SEQUENCE CONDITIOKAL PROSASILITY SLMS End State / Initiator Probability CD TRANS 8.0E-09 LOOP 4.9E-09 LOCA 8.2E 06 i total 8.3E-06 l ATWS j TRAK 5 0.0E+00 LOOP 0.0E+00 LOCA 0.0E+00 Tota 1 0.0E+00 SEQUENCE CONDITICkAL PRosAEILITIES (PROSASILITY ORDER) Sequence End State' Prob N Rec'* ' TZ toca -rt -afu NPI CD 8.2E-06 4.3E-01 j
- newrecovery credit for edited case i
SEQUENCE CONDITIONAL PROSABILITIE5 (SEQUENCE 00ER) SeqJence End State Prob N Rec ** /2 loca art afw HPI CD 8.2E 06 4.3E 01
- ruwrecovery credit for edited case Note: For movellabilities, conditional probability vetues are differentist vetues editch reflect the added risk chae to faltures associated with an event.'. Parenthetical values indicate a red.ction in '
' risk compared to a simiter period without the aristire feltres.
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PROSA81LITT FILE:- c:\\eap\\1999\\pwr.,,bst1. pro so Recovery Limit , I Event identifier: 266/92-010 i LER NO: 266/92-010 sa PRELBiINARY
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i , l - PRELBfINARY - i % j . BRANCH FREQUENCIES /PR08 ABILITIES i Branch System Non-tecov Opr felt j trans 6.4E-05 1.0E+00 ( loop 1.6E 05 ~ 3.M-01
Ioca 2.4E-06 4.3E-01 l rt 2.8E 04 1.2E-01 rt/toop 0.0E+00 1.0E+00 amers. power 2.9E-03 8.0E 01 , afW 3.8E-04 2.6E-01 l efm/emers. power 5.bE 02 3.4E-01 erfw 1.0E+00 7.0E-02 1.0E-03 pory.or.ory.chatt 4.0E 02 1.0E+00 - pory.or.arvoreseet 2.0E 02 1.1E-02 pory.or.srv. reseat /energ. power 2.0E-02 - .1.0E+00 Seet.Ioca 0.0E*00 1.0E+00 ep.rectst) 0.DE+00 1.M+00 i ep. rec 4.5E-01 1.0E+00 i hPI 1.0E-03 > 1.0E+00 S.4E 01 > 1.0E+00 l Branch Model: 1.0F.2 Train 1 Cond Prob: 1.DE 02 * foited ' Train 2 Cond Prob: 1.QE-01 > Failed NPI(r/B) 1.0E 03 > 1.0E+00 8.4E-01 = 1.0E+00 ' 1.0E-C2 Branch kodet: 1.0F.2+ ope Train 1 Cond Prob: 1.0E 02 * Felled Train 2 Cond Prob: 1.0E-01 > Falled hpr/ hpl 1.5E 04 '1.0E+00 1.0E-03 I pory.open 2.0E 02 1.0E+00 4.0E-04
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Event Identifier: 266/92-010 t , i ! . ! -, > . LER NO: 266/92-010 . B-5 PRELBHNARY. .
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9 t IF' ee a.non.e iE09t$C DsAC.'5B @w b.fil 9lC l l l wn-..-.. -.. - - = AB F*RA C'* - At 1450 on Dece=.ber 8,1992, while Point Esach Fuclear Flant (FENP) Units 1 and 2 were cperating at 100% and 954 power respectively, it was discovered that Inservice Tests IT-40, ' safety Injection Valves (Ratrtarly), Unit 1,* a.nd 1T-45, ' safety Injection valvss (Quarterly), Unit 2,* could lead to the isolation of all available flow paths for the ostety insection (s!) pumpe. Tests IT-40 and IT-45 perform quartarly stroke tasts of oafety injection / containment spray minimum flena recirculation line isolation valves 18251-857A and 1&2sI-8973 (hereinafter referred to as valves 497 ALE). IT-40 and IT-45 ploes the plant in a condition in which purp damage could occur if ths SI pumps cutcastically started while reactor coolant system (Res) presswa was greater than SI pusep shutof f head and either Vales 497A or Valve 4973 remained shut. Operating the SI pumps at shutof f head would cr use smap damage after approximately one mlnute. The torts were isst performed on 11/15/92 (Unit 1) and 11/19/92 (Unit 2).
A 4-hour 3DLC ENS notification was made in accordance with 10 CTR 50.72(b)(2)(iii)(D). The IUtc Resident 2n:peeter was also notified. A Proba.bilistic Risk Assessaant (FRA) was subsequently performed and determined that the probability of this event occurring is approximately 2.0 E-6 eTents/ year, or an increased pump damage risk of approximately 2 percent. Due to the increased risk of da.maging the SI pumps by testing Valvse 897AER on a quartarly frequency, the tests will subsequently t.e performed on a cold shutdown frequency.
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! ) . yvrNT DESCRIPTION At 1450 on December 8,1992, while point Beach Neelear Flaat (PENF) Units 1 and 2 were r operating et 100% and 95% power respectively, it was discovered that Inserviso Tests IT- -
40, ' Safety Injection valves (Quarterly), Unit 1,* and IT-45, " Safety Injection Valves i (Quarterly), Unit 2,* place the plant in a condition in which damage could eseur to both 81 pumps.
IT-40 and IT-45 perform quarterly stroke tests of SI/CS mini-rocirculation
line isolation valves 1&2sI-897A and 15251-8975 (hereinafter referred to as Valves
' 097A&B). The damage could occur if the SI pumps automatically started while either i Valve 497A or 8973 was shut.
Operating the SI pumps with either Valve 897A or 8973 shut would cause SI pump damage due to operation of the SI pumps at shutoff head without cinianas recirculation flow if reactor coolant system (RCS) pressure was greater than SZ ' pump shutoff head and plant operators failed to open one of the. valves, 897A or 8973, . ! trithin approximately one minute.
The tests were last performed on 11/15/92 (Unit 13 and 11/19/92 (Unit 2).
Upon identification of this condition on December 8, 1992, a d-hour l , NRC EMS notification was made in accordance with 10 CFR 50.72(b)(2)(iii)(D).
The NRC . Resident Inspector was also notified.
' Although the EMS notification identified that the containment spray (Cs) pumps could cleo be damaged under the same circumstances, this condition is now considered to be of loss concern.
The CS pumps have a flow path to containment regardless of the position .cf Valves 897AER unless both of the two parallel motor-operated discharge valves per CS ! pump fail to open on an automatic signal. Because the Cs pump discharge valves are powered from separate safeguards trains, concurrent f ailure of both pairs of discharge ! , valves is not a credible event.
f! A Probabilistic Risk Assessment (FRA) was subsequer.tly performed and determined that the probability of an automatic initiation of SI occurring while either Valve 897A or 8975 j is shut is approximately 1.0 E-6 events / year, or an increased pump damage risk of ' cpptcaimately 2 percent.
' Section XI of the ASME Soiler and Pressure vessel Code, !
Article IW-3412a,1986 Edition, allows plants to identify those valves which cannot be tasted during plant operation and provide for full-stroke testing of these specific j ' valves during cold' shutdowns. The PSNP In&ervice TestiW1 (IST) program acceemts for ' valves requiring this type of testing in Appendia 8, 'lotd Shutdown Justifications.* , Th3refore, due to the elevated risk of pump damage while testing Valves 897A&E during plant operation, testing of valves 897A&R will be deferred to periods when the s'ocpective unit is in cold shutdown and both SI pumps may be taken out of servloe.
, Dicconsin Electric addressed a related issue in a letter to the NRC dated July 24, 1985.
The letter, subnitted in accordance with 10 CFR 21, motified the NRC that the failure of a oingle component in the control circuitry for the SI recirevlation th isolation valves could, under specific circumstances, result in the failure of
safety in$cetion pumps. During a post-implementation review of the Emergency Operating . - Procedures (RoPs), it was discovered that a failure of the power supply areaker in the remote control circuitry for Valves 897A&R would result in those valves closing.
This i i failure would simultaneously result in loss of valve position indication and defeat the annunciation for 897A&R valve closure on the main control board. The corrective actions ' cpecified in response to this issue included gagging the manual handwheel operstars on Valves 897A&B in the open position so that the automatie operators would be overridden.
This corrective action was also referenced in our response to NRC IRS 86-03, ' potential Failure of Multiple ECCS Pumps Due to Single Failure of Air-operated valve La the - i . i!inimum Flow Recirculation Line," dated November 12, 1986. Sowever, quarterly inservice
- ctrche tests in whi-h the valves were ungagged for a short period of time ar,d
' repositioned for testing were considered at that time to pose no significar,t iners. ae in rick to the 21 pumps.
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u Point Beach Nuclear Plant, Unit l' s l8 {e l81812 I616 912 0 l1IO DID 011 nIA - - .m,o _ _.e _ _ _ _, _. , , BQUIPMENT DESCRIPTION (Note: Information in () indicates Energy Industry Identification systesa (EZIs) identifiers) ! An orificed minimum flow bypass line is provided at the discharge of each SI (39] pump (P) to recirculate flow to the refueling water storage tank (Rust)('TE] through a onesson l
he:dar (or, *sini-recirc* line) - in the event the pumps are run while the RC8 [AB) pressure is above the pumps' shutoff head. These bypass lines also permit the performance of periodic surveillance tests required by the Technical specifications to ? prove pump operability. The recirculation line is provided with air-operated Leolation valv's 897A&B (Isv), in series, which are closed to prevent flow of contaminated water e to the RWst when in the containment sump recirculation phase _ following an accident.- Because valves 897A&S fail shut, they are normally gagged open to prevent closure on a l loss of instrument air.
! will overheat and quickly deteriorate.If the SI pumps are operated without a flow path, the pumps ! . Valves 897A&E are interlocked with containment pump *B* isolation Valves 1&2sI-851A&B (15v) (hereinafter referred to as Valves 851A&B). These motor-operated gate valves are normally closed except when required for containment.eump recirculation following an
This interlock insures that Valves 851A&S cannot be opened until at least one
cecident.
valve, 897A or 8978, is closed which prevents the inadvertent release of containment i cump vapor or liquid to the RWst during the containment sump recirculation phase of
long-tem cooling following a design basis accident.
' , The manual handwheel operators on Valves 897AEB are currently maintained in the open position to prevent closure on a loss of instrument air.
' Mn1% The re-evaluation of the PBNP quarterly inservice testing practices for Valves 097A&E .
l was prompted by INFO Nuclear Network Message et 5692, *I,oss of All ICCs Pumps During
Monthly surveillance Testing," transmitted on November 24, 1992, by Calvert Cliffs i
I:uclear Plant.
Prior to this re-evaluation, quarterly inservice stroke tests in which valves 897A&R f were ungagged for a ahort period of time and repositioned for testing were deemed necessary and considered to pose no significant increase in risk to the SI pumps.
, , carArc?!vr actions A.
Immediate . 1.
Further testing of valves 897A&B was suspended.
B.
Short ters: 1.
A Probabilistic Risk Assessment (FRA) was performed to determine the probability of an SI actuation during the time Valves 897A&R are being tested.
station logs were reviewed and indicated that the approximate time to complete IT-40/45 is on the order of two hours (however, the valves are not shut for the full duration of the test) and therefore the time that the valves are ungagged each calendar quarter is small. Given this information, ' it was determiner. that the probability of this event occurring is . approximately 1.0 E-f, events / year, or an increased pump damage risk of ' approximately O getcono - -- - ww - --t-- - , . . .
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The pumps' sanufacturst, Byron Jacksca, was consulted and stated that the SI piutps can be operated at shutoff head for up to one minute before pump degradation begins. Therefore, control operators would have up to one minute after an automatic pump start to restore the flow path *if instrument air is available.
If instruament air is not available, the valves would require manual handwheel operation.
C.
Iong Tern 1.
A Cold shutdown Justification (CJJ) for Valves 1&2EI-897A&B will be included in the IST program to allow testing on a cold shutdown frequency. This change was substitted to the NRC on December 23,11r92.
2.
Test procedures will be devoleped to provide for the inservice testing (stroke time, fail-safe, position indication verification, leak rate testing) of Valves 1&252-897A&B on a cold shutdown frequency. This will be courplated by the Operations Group by February 28, 1993.
3.
The operations Group will ensure that all other valves currently tested under l Procedures IT-40 and IT-45 on a quarterly basis will continue to be tested on a quarterly basis. Procedures to aceceplish this testing will be developed . if riecessary. All necessary procedure revisions will be completed or new procedures developed by February 26, 1993.
i . PIPOFI A P " I*Y This event is being reported under the requirements of Id CFF. 50.73(a)(2) (v)(D), 'The licensee shall repert...any event or condition that alone could have prevented the fulfillment of the tafety function of structures or systees that are needed to mitigate th3 corsequences of an accident.* A 4-hour KRC EM5 notification was made in accordance with 10CFR50.72(b)(2)(iii)(D). The NRC Resident Inspector was also notified.
KArr*Y ASSESSMEN* A Probabilistic Risk Assessment (FRA) was perforsned and determined that the probability cf operating an 51 pump at shutoff head after Valves 897As3 have been ungagged and the valves have failnd shut is 1.0 E-6 events / year. The probability of failure of the KI l j pumpe for other reasons is calculated to be 5.3 E-5 events / year.
Therefore, this ) id:ntified condition would reruit in an increased risk of failure of the SI pumps of about 24.
This condition is a small contributor to the failure of the SI pumps. Eence, i ' th3 probability of pep damage occurring as a result of the scenario described above is d;tornined to not be a significant contributor to core damage frequency. The safety of th3 plant and the health and safety of the public and plant employees were not $eopardised by this plant condition.
. CY%TPf e IMPiTCATIONS No generic implications have been identified.
ITM?iAR occurArNers ,
Th:re have been no similar occurrences identified at PRKP.
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