ML14184B070

Forwards Draft ASP Evaluation for Four Events Which Occurred in 1992.Events Involved Inoperability of Safety Injection Pumps & Loss of Offsite Power

ML14184B070
Person / Time
Site:	Robinson
Issue date:	06/08/1993
From:	Mozafari B Office of Nuclear Reactor Regulation
To:	Dietz C Carolina Power & Light Co
References
NUDOCS 9306150097
Download: ML14184B070 (32)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 8, 1993 Docket No. 50-261 Mr. C. R. Dietz Vice President Carolina Power & Light Company H. B. Robinson Steam Electric Plant, Unit No. 2 Post Office Box 790 Hartsville, SC 29550-0790

SUBJECT:

PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSES FOR FOUR EVENTS AT THE H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 The NRC has developed preliminary accident sequence precursor (ASP) evaluations as a follow-up of significant Licensee Event Reports (LER) submitted in 1992. Enclosed are draft ASP evaluations for four events at the H. B. Robinson Steam Electric Plant, Unit No. 2, which occurred in 1992. The events involved inoperability of the safety injection pumps and a loss of offsite power.

Your review and comments on the analyses of these events would be appreciated.

In particular, comments are requested on the characterizations of possible plant response, given the event occurrence. We are also interested in comments concerning whether or not the individual analyses reasonably represent the plant safety equipment configurations and capabilities that existed at the time of the events. Comments on the analyst's assumptions regarding equipment recovery probabilities are also sought.

As discussed with Mr. Robert Prunty of your staff, we request that your comments be provided by June 25, 1993.

We will review your comments and revise the final ASP analyses, as appropriate. If you have questions regarding this matter, please contact me at (301) 504-2020.

This request is covered by the Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994. The estimated average number of burden hours is 80 person-hours per owner response, including the time required to assess the new recommendations, search data sources, gather and analyze the data, and prepare the required letters. Comments on the accuracy of this estimate and suggestions to reduce the burden may be directed to the Desk Officer, Office of Information and Regulatory Affairs (3150-0011),

00 1C%

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06150097 930608 pDR ADOCK 05000261

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___PDR

June 8, 1993 Mr. C. R. Dietz

- 2 NEOB-3019, Office of Management and Budget, Washington, DC 20503, and to the U.S. Nuclear Regulatory Commission, Information and Records Management Branch (MNBB 7714), Division of Information Support Services, Office of Information and Resources Management, Washington, DC 20555.

Sincerely, ORIGINAL SIGNED BY:

Brenda Mozafari, Project Manager Project Directorate II-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosure:

Preliminary ASP Analysis B.8 cc w/enclosure:

See next page DISTRIBUTION:

Docket File, NRC/Local PDRs PD II-1 Reading File S. Varga G. Lainas S. Bajwa B. Mozafari E. Jordan MNBB 3701 W. Beckner 10-E-4 P. Anderson OGC ACRS (10)

E. Merschoff, R-II cc:

Plant Service list OFFICE _L.

PE P,1D?:DRPE AD:PD21:DRPE NAME PAn son ozafari:dt SBajwa_

DATE 06/

/93 06/ (

/93 06/ 9( /93 Document Name:

ROBASP.LTR

Mr. C. R. Dietz H. B. Robinson Steam Electric Carolina Power & Light Company Plant, Unit No. 2 cc:

Mr. H. Ray Starling Mr. Dayne H. Brown, Director Manager - Legal Department Department of Environmental, Carolina Power & Light Company Health and Natural Resources Post Office Box 1551 Division of Radiation Protection Raleigh, North Carolina 27602 Post Office Box 27687 Raleigh, North Carolina 27611-7687 Mr. H. A. Cole Special Deputy Attorney General Mr. Robert P. Gruber State of North Carolina Executive Director Post Office Box 629 Public Staff -

NCUC Raleigh, North Carolina 27602 Post Office Box 29520 Raleigh, North Carolina 27626-0520 U.S. Nuclear Regulatory Commission Resident Inspector's Office Mr. Heyward G. Shealy, Chief H. B. Robinson Steam Electric Plant Bureau of Radiological Health Route 5, Box 413 South Carolina Department of Health Hartsville, South Carolina 29550 and Environmental Control 2600 Bull Street Regional Administrator, Region II Columbia, South Carolina 29201 U.S. Nuclear Regulatory Commission 101 Marietta St., N.W., Ste. 2900 Mr. H. W. Habermeyer, Jr.

Atlanta, Georgia 30323 Vice President Nuclear Services Department Mr. Ray H. Chambers, Jr.

Carolina Power & Light Company Plant Manager Post Office Box 1551 Carolina Power & Light Company Raleigh, North Carolina 27602 H. B. Robinson Steam Electric Plant Post Office Box 790 Hartsville, South Carolina 29550 Public Service Commission State of South Carolina Post Office Drawer 11649 Columbia, South Carolina 29211

ENCLOSURE PRELIMINARY B.8 LER Number 261/92-013, 261/92-014, 261/92-017, 261/92-018 Event

Description:

Both Safety Injection Pumps Out of Service and Loss of Offsite Power Date of Event:

June 16, 1992, through August 22, 1992 Plant:

H. B. Robinson, Unit 2 B.8.1 Summary Both safety injection (SI) pumps were out of service for 1.5 h on July 10, 1992, while H. B. Robinson was at 100% power. The "B" SI pump was rendered inoperable because plastic sheeting material obstructed the pump's recirculation line. The plastic material was used during a design modification during the refueling outage that ended on June 18, 1992. The "A" pump was out of service for 1.5 h on July 10, 1992, because of a blown control power fuse in the pump's breaker closing circuit. On August 22, 1992, with the plant operating at 100% power, the loss of the startup transformer resulted in loss of one of the two emergency buses and an instrument bus. Following a subsequent reactor/turbine trip, the transfer of the other emergency bus to offsite power failed and resulted in a total loss of offsite power (LOOP). Following the LOOP, on August 24, 1992, the "B" SI pump recirculation line was again found to be obstructed with the plastic sheeting material from the outage modification.

The conditional core damage probability for the 1.5 h that both SI pumps were inoperable (LERs 261/92 013 and -014) is 1.5 x 10-0, the conditional core damage probability for the time period when the "B" SI pump was inoperable (LERs 261/92-013 and -018) is 1.5 x 10-', and the conditional core damage probability for the LOOP event (LERs 261/92-013, -017, and -018) is 2.0 x 10-'. The sum of the conditional core damage probabilities for these three related events is 3.5 x 10-'.

The relative significance of these events compared to other postulated events at H. B. Robinson Unit 2 is shown in Fig. B.10.

B.8.2 Event Description On July 8, 1992, at 2307 hours0.0267 days <br />0.641 hours <br />0.00381 weeks <br />8.778135e-4 months <br />, the B SI pump was declared out of service because of low flow on the pump's recirculation line. Plastic sheet material was found in the "B" SI pump minimum flow line.

The plastic material was believed to be from a purge dam that had been fabricated for welding operations for a modification to the minimum flow line for the residual heat removal (RHR) system during the cycle 14 refueling outage. The refueling outage ended on June 18, 1992. It is believed the material was introduced as a result of breakage of one of the 9-in.-diameter purge dam pieces. A small portion of the material was introduced into the RHR system, the refueling water storage tank (RWST), and SI and containment spray (CS) pump suction piping. The debris was removed through system flushing.

LER NO: 261/92413, 014, 017, 018 B-39 PRELIMINARY

PRELIMINARY L

26182.15 417 A018 LZR 2612415 414 LM 26182 f4, 4w47*411I 1M7 124 s2-5

'tl I

3U0"+1 IMAFW Fig. B.10.

Relative event significance of LERs 261/92-013, -014, -017, and -018 compared with other H. B. Robinson 2 potential events.

On July 9, 1992, at 1839 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />6.997395e-4 months <br />, with the plant still at 100% power, an attempt was made to start the "A" SI pump. During this attempt, one of the two control power fuses in the pump's breaker closing circuit blew. The fuses were replaced, and the pump was returned to service 1.5 h later, at 2009 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.644245e-4 months <br /> on July 9, 1992.

At 2030 hours0.0235 days <br />0.564 hours <br />0.00336 weeks <br />7.72415e-4 months <br />, on July 9, 1992, a plant shutdown to the hot shutdown condition was initiated because of the continued inoperability of the "B" SI pump. On July 12, 1992, at 0812 hours0.0094 days <br />0.226 hours <br />0.00134 weeks <br />3.08966e-4 months <br />, the "B" SI pump was returned to service following repeated flushing of the SI system.

Operability tests were also performed for the RHR and CS systems. The plant returned to service on July 12, 1992.

On August 22, 1992, with the plant at 100% power, a LOOP occurred at 1007 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.831635e-4 months <br /> because of the loss of the startup transformer. The loss of the startup transformer caused a loss of emergency bus E-2 and instrument bus 4, and a turbine runback.

The "B" emergency diesel generator (EDG) started and supplied emergency bus E-2. At 1009 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.839245e-4 months <br />, the turbine and reactor tripped on high steam generator level. At 1010 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.84305e-4 months <br /> the auxiliary transformer tried to transfer its loads to the startup transformer but failed because the startup transformer was not operational. Iis resulted in a LOOP to the other emergency bus (E-1). The "A" EDG started and supplied emergency bus E-1. A manual SI was initiated at 1018 hours0.0118 days <br />0.283 hours <br />0.00168 weeks <br />3.87349e-4 months <br /> because the pressurizer level had fallen to less than 10% during the initial transient. At 1037 hours0.012 days <br />0.288 hours <br />0.00171 weeks <br />3.945785e-4 months <br /> the manual SI was terminated. At 1103 hours0.0128 days <br />0.306 hours <br />0.00182 weeks <br />4.196915e-4 months <br /> natural circulation was verified, with RCS temperatures stabilized at 500F. Repairs to the startup transformer were completed and normal power alignment restored to the emergency busses at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> on August 23, 1992.

On August 24, 1992, following the LOOP and before plant restart, the "B" SI pump was tested and declared inoperable because of low flow in the recirculation line. The "A" SI pump was also declared inoperable because of reduced flow in its recirculation line. Investigation revealed that additional plastic sheeting material from the RHR system modification performed during the cycle 14 refueling outage had partially blocked the "B" SI pump recirculation line. No debris was found in the *A" pump recirculation LER NO: 261/92-013, 014, 017, 018 B-40 PRELIMINARY

PRELIMINARY line, and the flow was within the required limits. Therefore, the "A" line was considered to have been operable throughout the event.

B.8.3 Additional Event-Related Information H. B. Robinson has two RHR pumps, which take suction from the RWST or the containment sump. The system can discharge to the reactor coolant system (RCS) cold legs or to the suction of the SI and CS system pumps. The RHR pyimp recirculation lines run back to the suction of the pumps.

The SI system uses two pumps that can take suction from the RWST or the RHR pump discharge.

Each pump has a recirculation line to provide pump cooling. The recirculation lines return to the RWST. The RHR, SI, and CS pumps all share a common suction line from the RWST. The original SI system included three pumps; however, one of the pumps has been removed from service for an extended period of time.

During power operation the main generator supplies 4160-Vac buses I and 4 via the unit auxiliary transformer (UAT) (see Fig. B.11). Buses 2 and 5 are also supplied from the UAT via buses I and 4, respectively. Bus 3 is supplied from offsite power via the startup transformer (SUT). Emergency bus E-1 is supplied from the main generator via the UAT, bus I and bus 2. Emergency bus E-2 is supplied from offsite power via the SUT and bus 3. Upon loss of the main generator, the UAT transfers all loads to the SUT. If this transfer fails, the emergency buses are isolated from the non-safety-related buses and the EDGs start and load onto the buses.

B.8.4 Modeling Assumptions These four licensee event reports (LERs) are analyzed together because of the unavailability of the "B" SI pump throughout the entire time period. The root cause of the "B" SI pump inoperability was the plastic sheeting material from the RHR system modification performed during the cycle 14 refueling outage.

The first event was modeled assuming that the two SI pumps were inoperable for 1.5 h. For the second event, it was also assumed that the "B" SI pump was inoperable from the time the plant went critical following the completion of the plant outage on June 17, 1992 (64.5 d), until the second recirculation line plugging event occurred onJuly 24, 1992.

The LOOP event was modeled as plant-centered. Probabilities for LOOP nonrecovery (short term) and failure to recover ac power prior to battery depletion were revised to reflect values associated with a plant-centered LOOP (see ORNLINRC/LTR-89/1 1, Revised LOOP Recovery and PWR Seal LOCA Models, August 1989).

B.8.5 Analysis Results Tle conditional core damage probability for the 1.5 h that both SI pumps were inoperable (LERs 261/92 013 and -014) is 1.5 x 10-6. The dominant core damage sequence for this event, shown in Fig. B.12, LER NO: 261/92-013, 014, 017, 018 B-41 PRELIMINARY

PRELIMINARY involves a LOCA followed by a failure of high-pressure injection.

The conditional core damage probability for the time period when the "B" SI pump was inoperable (LERs 261/92-013 and -018) is 1.5 x 10-'. The dominant core damage sequence for this event, shown in Fig. B. 13, involves a LOCA followed by a failure of high-pressure injection. The conditional core damage probability for the LOOP event (LER 261/92-017) is 2.0 x 10'. The dominant core damage sequence for this event, shown in Fig. B. 14, involves failure to restore emergency power and a resulting reactor coolant pump seal LOCA.

The sum of the conditional core damage probabilities for these three related events is 3.5 x 10-'.

FROM rRom I ISKV S~TCYAED 230KV SWrFCW4ARD Y FMOM 230KV sWrCurAND MM.

N H

GENEATOR SUF1-2f z

UAT-2 NS-2 (24(S~IF~

s DS T5 I

ma 7Y-T

. aa p

44 me a

wa g2 m e1 a L. a be ss ? a as 1e2 TO 4US Et TO IUS (2

Fig. B.11.

H.B. Robinson electrical distribution system.

LER NO: 261/92-013, 014, 017, 018 B-42 PRELIII ARY

0 S

PRELIMINARY LOCA RT Afw POWeM~

LPEN SEQ END No2. STATE OK 71 CD 72 CD OK 73 Co 74 CD OK 75 CD 76 co 77 co 78 ATWS Fig. B.12.

Dominant core damage sequence for LERs 261/92-013 and 014.

CA M

ww HM PAI OK 71 CD 72 Co OK 73 co 74 CD OK 75 CD 76 CD 77 Co 78 ATWS Fig. B.13.

Dominant core damage sequence for LERs 261/92-013 and 01S.

LER NO: 261/92-413, 014, 017, 018 B43 PRELIMINARY

PRELIMINARY RT/

PORV/

PORV/

SEAL EP REC Hp I

RV OOP LOOP EP AFW RESAT LOCA (LONC)

OPEN STATE OK OK 41 Co 42 CD OK OK 43 Co 47 CD 45 CD OK 46 CD 47 CD 48 Co OK 49 CD 50 CD OK Sl CD 52 CD 53 CD OK 54 Co 55 Co 40 ATWS Fig. B.14.

Dominant core damage sequence for LERs 261/92-013, 017, and 018.

LER NO: 261/92-013, 014, 017, 018 B-"

PRELIMINARY

PRELIMINARY CONDITIONAL CORE DAMAGE PROBABILITY CALCULATIONS Event Identifier: 261/92-013, 014 Event Descrfption: BoticSI pus* 005 at Robinson for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Event Date:

07/10/92 Plant:

Point Beach 1 UNAVAILABILITY, ORATION-1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> NOM*RECOVERAsLE INITIATIG EVENT PROBABILITIES TRANS 9.6E*05 LOOP 8.8E-06 LOA1.5E-06 SEQUENCE CONDITIMAL PROBABILITY SLAIS.

End State/Initiator Probabflity TRANS 1.5E-09 LOOP 9.3E-10 Total 1.S-06 Tots(1.5E-06 TRANS 0.0E00 LOOP LOCA 0.0E+0 SEQUENCE CONDTIONAL PReoASILIltS (PROSAILITY ORDER) seqisnce on State Prob te" 72 tece *r*fw IIPI CD 1.3x-06 3.dE-01 7t loae *rit -ft -UPI UPI-p 2.5E 07 A.9E-*0 ran-rcovswewy credit for edited case SEQUENCE CONITIONAL PABILITIES (SEQUENCE COAER)

Sow"*

End state Prob

• t ec**

71 toce *Ft afe IWI P/-HPI CD 2.5E-0'7 6.9E-02 72 teca rt **ft iPI CD 13E-06 3dE-01

• • non-recovery credit for edited cse mote: for stwevlabilitias, cinditfonal probabitfty values are dtfffrmntist vautes,A ch ref lect the added rfak the to faiur aesociatal afth an vent. Parenthetict values indicate a redutor in risk compared to a stt piefd without the existing falttes, SEQUENCE NIL C lasppest tpurbseet.cop GRANCH MODEL.

CAaprsadet*1ptheacht.stt PtaBABILITY FILE,.

C*\\*sppreveodasl.purbet 1.pro Event Identifiers 261/92-013, 014 LER NO: 261/92-013, 014, 017, 018 B45 PRELIMINARY

PRELIMINARY No Recover* Limit BRACH FREDlANCIEP/PRGSAILIT1ES:

,Branch systm' man-Recov irFaitI tram 6.E605, 1.OE.OO.

loocp 1.6E.053.-0 loca

~2.4.-06.

310 rt

~2.81-041.-0 rtlop 0.0E+00

.1.01.00.

Semrg-poer 2.9E-03

.8.0-01 efw 3.3E-04 2.6E-01 afmwae....power 5.05-02 3.4-1-C ofm 1.05400 7.01-02 porv.*rs.v.chstl 4.05-02 1.0e+00 pom-.or.ary.teseet Z.OE-02

.1-02 PorV.or.5rv-,reeemt/@MsGl.pomr 2.0E-02 1.01400

.eat.oca

.05400 1.06.00 epatec(sL) 0.05.00 1.05.00 ep. rw 4.55-011..0 Branch 105.00 1.0f.21 Tan1Cosnd Prob:

1.05-02 >Failed Trin Cond Probj t.0E-01 > Faited Trin I om' Probir I.06-02 Faited.

Train 2 taen, 1.55 S-02 3k Fated be bnh fds fI *er LER NO: 261/924013, 014, 017, 018 B-46 PRELMINARY

PRELIMINARY CONDITIONAL CORE DAMAGE PROBABILITY CALCULATIONS Event Identifier: 261/92-013, 018 Event

Description:

-B-SI pump 006 at Robinson Event Date:

06/18/92 - 08/25/92 Plant:

Point Beach I UMAVAILABILITY IURATION 1545.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> NON-RECOVERABLE INITIATING EVENT PROAABILITIES TRANS 9.9E-02 LOOP 9.1E-03 LOCA 1.61-05 SEQUENCE CONDITIONAL PROBABILITY SUNS End State/Initator Probablity CD TRANS 1.4E*07 LOOP 9.1E*08 LOCA 1.5t-04 Tota.5-0 ATUS TRANS 0.0E+00 LOOP 0.GE+00 LOCA 0.0E+08 Total 0.E+00 SEUCE CONITIONAL PROBABILITIES (PROBABILITY ORDER)

Seaer End State Prb N tec" 72 tece -rt **fv NPI CD 1.3E-4 3.6E*01 71 toca -rt -*fm *API IIPt/-P'S 2..t*05 4.2E-0t

• nerreaery credit for editel can SEQUENCE CONDITIONAL PROBABILITIES (SEQUENCE OADER)

Seance End State Prob N Rae*

71 (ac -rt *fu WI 1Ia-NP) 2.t-05 4.2t-Of 72 tece *Ft -af.

Bpi CD 1.31*4L 01

• • non*rcewry cradit for oedftd cass Kote: for awtitablities, toniditional probabitity values are differential vaues which reflect the added risk due to fSitures associated with an event.

Parenthetical values indicate a whation in risk copred to a simiter peris althout the existing falLures.

SEQUENCE NODEL:

c:1asppre das tspurbseat.c.

BRANCH MODEL:

eslaspprltodmts~ptbeecht.sa I PROgABILITY FILEs c:1aspprsmadelstprbott.pro Event Identiffer 261/92-013, 010 LER NO: 261/92-013, 014, 017, 018 B47 PRELIMINARY

PRELIMINARY No Recovery-Lilit BRANCH FREQUENCIES/PROBAI LITIES Branch System Non-Recoy Opr Fet trans 6.4E-o0 1.01+00 toP 1.6E-05 3.6E01 Loca 2.4E06 4.31E-01 rt 2.8E-04 121E01 rt/tooP 0.0E+00 1.0E+00 fmg.pmoer 2.9E-05 8.0E-01

• ft 3.81-04 2.6E,01 afm/eerg.pouer 5.0E-02 3.4E*01 efu 1.01.00 7.0E-02 pory.or.srv.chatI 4.0E02 1.0E+00 pory.orar.reet 2.01-02 1.1E*02 porv.or.srv.reseat/emerg.power 2.0E-02 1.0E+00 seatoce 0.0E+00 1.0E+00 lp.rectaI) 0.0E+00 1.0E+00 ep.rec 4.5E-01 1.01E+00 MPI 1.0E*03 > 1.01-01 8.4E-01 Branch Nodel: 1.DF.2 Train I Cord Prob.

1.0E-02 > Fated Train 2 Cand Prob:

1.0E-01 NPI(U/3) 1.0E-03 >, 1.01-01 8.6E-01 1.0E-02 Branch Model:

1.OF.2+epr Train I Cond Prob:

1.0E-02 > Faited Train 2 Cond Probg 1.0E-01 MfR/NP 1.5E-4 > 1.5E-02 1.0E+00 I.0E-03 Branch Model 1.0F.2*apr Trafn I Cand Probs 1.0E*02 > ftated Traina 2 Cmf Prob:

1.5E-02 pory.open 1.0E-02 1.0E+00 4.0E04

• branch mdel fite

• forced Event Identiffer: 261/92-013, 018 LER NO: 261/92413, 014, 017, 018 B-48 PRELIMINARY

PRELIMINARY CONDITIONAL CORE DAMAGE PROBABILITY CALCULATIONS Event Identifier: 261/92-013, 017, 018 Event

Description:

LOOP with 3B" SI Pap 00S at Robinson Event Date:

08/22/92 Ptant:

Point Beach 1 INITIATING EVENT MON*RECOVERABLE INITIATING EVENT PROBABILITIES LOOP S.0E-01 EQUENCE CONDITIONAL PROBABILITY SmIS End State/tnitiator Probabftity CD LOOP 2.OE-04 Total 2.0E-04 ATUS LOOP 0.0E+00 Total 0.01+00 SEQUENCE CONITIONAL PROBABILITIES (PROBABILITY ORDER)

Seqmne End State Prob N Rew 53 LOOP -rt/toop ere.pouer -aft/emrg.powr -pvmor.s.chatt CD 1.26-04 4.0E-01 SEAL.LOCA EP.UEC(SL) 54 LOOP -rt/toop emet.pouew *efatemr.pmier *poryor.ry.chatt 4.01 SALLOCA EP.itEC 55 LOOP *rt/toop arg.power afk/amerg.pomr CD*G-S t.I4E*01 52 LOOP -rt/top meg.power **f earg.poe *pory.or.srv.dchat C

1.1*5 3.35-01 SEAL LOCA -EP.RECCSL)

MPI 48 LOP *rtop rf p

.- het C

.40E*0 por~orsryrest/srs~earSEA.LOCA EP.REC(8) 45 LOOP -rt/loop *mr.poaw f

OPI(F/)C 4.*61 1.E-01 Snonrecovery credit for edited case SEQUENCE CODrIGMAL PROBMILITIE (SEauENCE OAER)

End State Preb t Re*

45 LOOP -rt/top -mrg.pam Of WP)(F/9)

C 4.46*e6 1.t-1 4

LOOP ret/op mrpor fers.pouer porV.orwary.chatt -

D 4.R'88 4.0E*01 pory.er~e 4tametmrg.pomr SEAL.LOCA EP.EC(SL) 52 LOOP *rtitep erg.our

-aft/eelrg.pouar -porv.er.srv.,chalt W

A.tE5-3.3E-01 SEAL.LOCA *EP.REC(SL)

HP 53 LOOP -ft/toop merg.powr -efu/erg.power *por.or.srvchalt CD 1.2E-04 6.0E-01 SEAL.LOC EP.RECCSL) 4 LOOP -rtoop mergepome afrme.power -porv.or.s.chatt

• 3.dE-5 4.0E-01 SEAL.LOCA SP.REC 55 LOP -rt/toop emg.power afu/merg.power C

1.9E'05 1.4E-01 Event Identifier: 261/92-013, 017, 018 LER NO: 261/92-013, 014, 017, 018

.49 PRELIMINARY

PRELIMINARY

• no-recovery cIdit for edited case SEUENCE NOEL:

C:\\asgara\\modetspwrbseest.cap BRANCH OEL:

C:\\asppra\\mdets\\ptechtI.sk1 PROBABILITY FILE:

C:\\spprakodets\\pueball.pro No Recovery Limit BRANCH FREUENCIES/PROBABILITIES Branck System Nan*Reov Opr Fall trans 6.*-S 1.0+0 LOOP 1.6E-05 1.6E-05 Brnch Nodat IMITOR Initiator Freq:

1.6E*0S loco 2.4E-06 4.3E-01 ft 2.8E-04 1.2E*01 rt/toop 0.0E00 1.0E+00 emeg.power 2.9E-03 8.0E-01

• ON 3.8E-04 2.6E*01 afu/amrg.power 5.0102 3.4E-01

• t 1.0E+00 7.0E-02 poy.or.ry.chett 4.0E-02 1.0E+00 porvor.ryMreseet 2.0E-02 1.1E-02 porv.or.*rv.rwsett/ert.poer 20E-02 I 1.00 SEAL.LOCA 0.0E+00 > 2.3E*01 1.0E+00 Brtach Nodtt 1.0F.1 Trafe 1 Cw Probs 0.0+0* > 2.3E-01 E~~atEEKSL) 0.06+00 4.8E*0

.80 St~:tanh ied 1.0#K1 Tf I C nrobs 0.0E+00 ) 4.8E-01 E.......

4.5"*0

> 4.3E-021.40 Branck Noadet"l I.0F.1 Traf 1 Cn Probs 4.SE*01 4.3E-0 M

1.0E-05 1.01-0 8.4E*01 WrnhModet 1.0f.2 I CudProb 1.0*02 Faited Tran. 2 ConPob.

1.0E-01 (F/)

1.E03 I0E01

.E*02 Train 1 Condrtb 1.OE-02 > Faited Treta 2 Cnd Pb 1.0,-01 AP/M1.5-64>5t.02 1,04+0 1I0003 Train 3 etrbs

.56-0

• b~~

~tft~....

perv.

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• forced Event Identiffert 261/92-013, 014, 017, 018 LER NO: 261/92-013, 014, 017, 018 B-so PRELIMINARY

PRELIMINARY LICENSEE EVENT REPORT (LER)

FACILITY NAME: H. B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. DOCKET NO: 261 TITLE:

PLANT SHUTDOWN DUE TO SAFETY INJECTION PUMP INOPERABILITY EVENT DATE: 07/09/92 LER #: 92-013-00 REPORT DATE: 07/27/92 OTHER FACILITIES INVOLVED:

DOCKET NO: 05000 OPERATING MODE: N POWER LEVEL: 100 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR SECTION:

50.73(a)(2Xi)

LICENSEE CONTACT FOR THIS LER:

DAVID

CROOK, SR.

SPECIALIST, REGULATORY COMPLIANCE TELEPHONE: (803) 383-1179 COMPONENT FAILURE DESCRIPTION:

CAUSE: SYSTEM:

COMPONENT:

MANUFACTURER:

REPORTABLE NPRDS:

SUPPLEMENTAL REPORT EXPECTED: NO ABSTRACT:

On July 8, 1992, at 2307 hours0.0267 days <br />0.641 hours <br />0.00381 weeks <br />8.778135e-4 months <br />, H. B. Robinson Unit No. 2 entered a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Limiting Condition for Operation (LCO) due to inadequate recirculation flow for "B" Safety Injection Pump. An investigation of the cause of the low flow condition was initiated. At 2030 hours0.0235 days <br />0.564 hours <br />0.00336 weeks <br />7.72415e-4 months <br /> on July 9, 1992, a plant shutdown to hot shutdown condition was initiated.

The cause of this event is attributed to personnel error. Event investigation identified the cause of the "B" Safety Injection pump's reduced recirculation flow to be foreign material blockage within the associated minimum flow recirculation check valve and flow orifice.

This foreign material was subsequently identified as a plastic sheet material fabricated for use as purge dam material for welding operations associated with a recent modification to the RHR minimum flow recirculation system.

Removal of the debris was accomplished through extensive system flushing. Repairs associated with the "B" Safety Injection pump were satisfactorily completed at 0812 hours0.0094 days <br />0.226 hours <br />0.00134 weeks <br />3.08966e-4 months <br /> on July 12, 1992, and the plant was returned to service at 1301 hours0.0151 days <br />0.361 hours <br />0.00215 weeks <br />4.950305e-4 months <br />.

This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(A) as the completion of a plant shutdown required by the plant's Technical Specifications.

LER NO: 261/92413, 014, 017, 018 B-51 PRELIMINARY

PRELIMINARY I. DESCRIPTION OF EVENT On July 8, 1992, af2307 hours, H. B. Robinson Unit No. 2' entered a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Limiting Condition for Operation (LCO) due to inadequate recirculation flow for "B" Safety Injection Pump. An investigation of the cause of the low flow condition was initiated. At 2030 hours0.0235 days <br />0.564 hours <br />0.00336 weeks <br />7.72415e-4 months <br /> on July 9, 1992, a plant shutdown to hot shutdown condition was initiated. The NRC was notified of this shutdown via the ENS as required by 10 CFR 50.72(b)(1)(i)(A).

Following an additional day of investigation, it was determined that repairs could not be made within the allowed LCO time period. Technical Specification 3.3.1.2 requires that if the system cannot be restored within an additional forty eight hours of achieving hot shutdown condition, the unit must be placed in cold shutdown condition using normal plant cooldown procedures. This LCO would expire on July 11, 1992 at 2259 hours0.0261 days <br />0.628 hours <br />0.00374 weeks <br />8.595495e-4 months <br />. On July 11, at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, the licensee contacted the NRC to request a Regional Waiver of Compliance that would extend the period of hot shutdown condition from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.2 Following this discussion, NRC-Region II verbally granted the requested waiver, effective until July 13, 1992, at 2259 hours0.0261 days <br />0.628 hours <br />0.00374 weeks <br />8.595495e-4 months <br />.

Repairs associated with the "B" Safety Injection pump were satisfactorily completed at 0812 hours0.0094 days <br />0.226 hours <br />0.00134 weeks <br />3.08966e-4 months <br /> on July 12, 1992 and the plant was returned to service at 1301 hours0.0151 days <br />0.361 hours <br />0.00215 weeks <br />4.950305e-4 months <br />.

II. CAUSE OF EVENT The cause of this event is attributed to personnel error. Event investigation' has identified the cause of the "B" Safety Injection pump's reduced recirculation flow to be the result of foreign material blockage within the associated minimum flow recirculation check valve and flow orifice. This foreign material was subsequently identified as a plastic sheet material which had been fabricated for use as a purge dam material for welding operations associated with a recent modification to the Residual Heat Removal (RHR) minimum flow recirculation system. It is believed that the material was introduced as a result of breakage of one of four, nine inch diameter purge dam pieces.

The investigation identified that use of the plastic purge dams was abandoned after the attempted use of two dams was terminated by their removal from the RHR system piping because the plastic dams could not be adequately sealed. A small, unidentified portion of this material was inadvertently introduced into the system piping associated with the RHR system, the Refueling Water Storage Tank, and the Safety Injection and Containment Spray Pump suction piping.

H. B. Robinson Steam Electric Plant Unit No. 2, is a Pressurized Water Reactor in commercial operation since March, 1971.

2 H. B. Robinson Serial No. RNPD/92-1882, dated July 11, 1992.

' Adverse Condition Reports ACR 92-249 & ACR 92-250 LER NO: 261/92413, 014, 017, 018 B-52 PRELIMINARY

PRELIMINARY III. ANALYSIS OF EVENT At the time of this condition, all ECCS systems were operable with the exception of the "B" Safety Injection pump. With the plant at Hot Shutdown, the boron concentration was raised to cold shutdown levels to compensate for a steam line break accident, and licensee operators were reminded of the Emergency Operating Procedure Function Restoration Procedures that would mitigate an accident, should one occur with the loss of Safety Injection. Therefore, the Safety Injection Pumps were not an immediate concern to prevent a restart accident during a steam line break cooldown. The Charging Pumps were maintained fully operable as a backup to the Safety Injection Pumps. The amount of decay heat inventory was evaluated based on the Units' operation prior to shutdown, and it was determined that a single Charging Pump had capacity that exceeded the heat removal requirements. Additional operator attention to the capability of the Function Restoration Procedures would ensure a reliable compensatory performance could be achieved.

The basis of Technical Specification 3.3 states that "For a single component to become inoperable does not negate the ability of the system to perform its function, but reduces the redundancy provided in the system design and thereby limits the ability to tolerate additional equipment failures." The reactor had been placed in a hot shutdown condition at the time, borated to cold shutdown levels, and the decay heat from the fuel continued to decrease during the additional time repairs were being performed.

Additionally, a Probability Risk Assessment of the additional risk associated with the additional 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> extension requested was conducted by the licensees Nuclear Engineering Department, and found to be negligible.

Since the plant was borated to cold shutdown boron concentration and the Charging System was capable of providing adequate core cooling at the reduced heat loading, any reduction of margin created by one inoperable Safety Injection Pump had been compensated for.

IV. -CORRECTIVE ACTIONS Removal of the debris was accomplished through extensive system flushing. The SI system was operated at design flow rates, with no additional blockage of the orifice flow due to material present in that system.

Because of the plastic material geometry, it is believed that any material introduced into the Refueling Water Storage Tank would have settled to the bottom of the tank. It is unlikely for the material to be caught in the flow stream due to the geometry of the material and the relationship of the tank to the Safety Injection System's supply line. Therefore it was considered not to represent a blockage threat to any related equipment and piping systems.

The "A" SI Pump-had been operated at full flow following the completion of the RHR minimum flow recirculation modification, and has operated greater than thirty minutes in the minimum flow configuration with no evidence of foreign material blockage in that system. Additionally, flow testing was completed on both Containment Spray Pumps in a minimum flow configuration with acceptable results. These pumps are normally aligned with the minimum flow recirculation lines closed, with the pump discharge aligned directly to the containment.

LER NO: 261/92-013, 014, 017, 018 B-53 PRELIMINARY

PRELIMINARY This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(A) as the completion of a plant shutdown required by the plant's Technical Specifications.

V.

ADDMONAL INFORMATION A. Component Failures None B.

Previous Similar Events None LER NO: 261/92-013, 014, 017, 018 B-54 PRELIMINARY

PRELIMINARY LICENSEE EVENT REPORT (LER)

FACILITY NAME: H. B. ROBINSON UNIT NO. 2 DOCKET NO: 261 TITLE:

ENTRY INTO TECHNICAL SPECIFICATION 3.0 DUE TO SAFETY INJECTION PUMP INOPERABILITY EVENT DATE: 07/09/92 -

LER #: 92-014-00 REPORT DATE: 08/08/92 OTHER FACILITIES INVOLVED:

DOCKET NO: 05000 OPERATING MODE: N POWER LEVEL: 100 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR SECTION:

50.73(a)(2)(i)

LICENSEE CONTACT FOR THIS LER:

DAVID CROOK, SENIOR SPECIALIST REGULATORY COMPLIANCE TELEPHONE: (803) 383-1179 COMPONENT FAILURE DESCRIPTION:

CAUSE:

SYSTEM:

COMPONENT:

MANUFACTURER:

REPORTABLE NPRDS:

SUPPLEMENTAL REPORT EXPECTED: NO ABSTRACT:

On July 9, 1992, H. B. Robinson Unit No. 2 was operating at one hundred percent power. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Limiting Condition for Operation (LCO) was in effect in accordance with Technical Specification 3.3.1.2.b for the "B" High Head Safety Injection (SI) Pump due to unscheduled maintenance. At 1839 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />6.997395e-4 months <br />, while starting "A" High Head SI Pump to verify flow measuring equipment operation, one of two control power fuses blew in the pump breaker closing circuit, and licensee operators declared the "A" SI Pump inoperable. Due to the inoperability of all High Head Safety Injection pumps, the action statement for Technical Specification 3.0 was entered.

Both control power fuses were removed from the "A" SI Pump breaker and replaced with identical fuses from the "B" SI Pump breaker. At 2009 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.644245e-4 months <br />, after three successful pump starts from the Control Room, the "A" SI Pump was declared operable, and the action statement for Technical Specification 3.0 was exited. Two possible causes have been identified for the fuse failure. Either the fuse failed to withstand its tested, nominal breaker closing current under the fuse's closing curves, or there occurred a current of enough magnitude and duration to blow the fuse during this one closing.

This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(B).

LER NO: 261/92-013, 014, 017, 018 B-55 PRELIMINARY

PRELIMINARY I. DESCRIPTION OF EVENT On July 9, 1992, H, B. Robinson Unit No. 2' was operating at one hundred percent power. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Limiting Condition for Operation (LCO) was in effect in accordance with Technical Specification 3.3.1.2.b for the "B" High Head Safety Injection (SI) Pump due to unscheduled maintenance.2 At 1839 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />6.997395e-4 months <br />, while starting "A" High Head SI Pump to verify flow measuring equipment operation, one of two control power fuses blew in the pump breaker closing circuite, and licensee operators declared the "A" SI Pump inoperable. Due to the inoperability of all High Head Safety Injection pumps, the action statement for Technical Specification 3.0 was entered. This action requires that, if a Limiting Condition for Operation cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in hot shutdown within eight hours, and in cold shutdown within the next thirty hours, unless corrective measures are taken that permit operation under the permissible Limiting Condition for Operation statements for the specified time interval as measured from initial discovery.

The NRC was notified of the entry into the Technical Specification action statement via the ENS on July 9, 1992, at 1927 hours0.0223 days <br />0.535 hours <br />0.00319 weeks <br />7.332235e-4 months <br /> pursuant to 10 CFR 50.72(b)(1)(ii).

Both control power fuses were removed from the "A" SI Pump breaker and replaced with identical fuses from the "B" SI Pump breaker. At 2009 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.644245e-4 months <br />, after three successful pump starts from the Control Room, the "A" SI Pump was declared operable, and the action statement for Technical Specification 3.0 was exited.

II. CAUSE OF EVENT Although the root cause of this event cannot be specifically determined, two possible causal factors have been identified. The manufacturer concluded the fuse was progressively weakened by repeated breaker closures until it opened to clear the circuit. Although it is presumed the fuse performed as designed, the first possible cause is a failure of the fuse to withstand the tested and nominal breaker closing currents under the fuse's published curves.

le second possible cause is that a current anomaly occurred with a current of enough magnitude and duration to blow the fuse during this one closing cycle that did not occur during previous or subsequent closings.'

'H. B. Robinson Steam Electric Plant Unit No. 2, is a Westinghouse Pressurized Water Reactor in commercial operation since March, 1971.

2 LER 92-013, Plant Shutdown Due to Safety Injection Pump Inoperabiity 3 Westinghouse Type DB-50

'EIS Codes System: BQ; Component: CKTBKR; Manufacturer: W120, B569 LER NO: 261/92-013, 014, 017, 018 B-56 PRELIMINARY

PRELIMINARY III. ANALYSIS OF EVENT Entry into Technical Specification 3.0 represents a "condition prohibited by the plant's Technical Specifications." Therefore, this LER is submitted pursuant to the requirements of I CFR 50.73(a)(2)(i)(B).

The safety significance of this condition is considered to be minimal. At the time of this condition, all ECCS stems were operable with the exception of the "B" Safety Injection Due to the relatively short period of time that both were inoperable, the likelihood of plant transient requiring safety injection time period is considered to be negligible. In addition, Function Restoration Procedure FRP-C.1 provides plant operators with actions to restore core cooling available if Safety Injection flow in all trains is not obtained.

IV. CORRECTIVE ACTIONS An investigation was initiated to determine the cause of the fuse failure.' The blown fuse was installed in this circuit on April 18, 1992 under Work Request WR/JO 91-AGNY1, replacing a Bussmann REN-10 fuse. Calculation No. RNP-E-9.005, performed under the H. B. Robinson Fuse Control Program, verified the adequacy of the fuse for this application.

On July 10, 1992, as part of the investigation, licensee engineers recorded the closing circuit current draw during closing of the breaker. The results demonstrated that the recorded value was 11.55 peak amperes during the 156ms closing cycle, which falls within the breaker manufacturer's nominal values.

Time-current curves for the control power fuse indicates it could withstand up to 55 amperes for 150ms, which is two and one half times the manufacturers' nominal rating five times the measured current draw on the DB-50 closing circuit. Additionally, the fuse can withstand 15 amperes fur five minutes, or 20 amperes fur 50 seconds. The time-current curves indicate the fuse is adequate or the requirements of the breaker (when compared to the manufacturers nominal time-current values and CP&L tested values) and should be capable of withstanding repeated closing operations. This fuse is presently being used in DB-50 closing circuits at H. B. Robinson and there have been no other reported incidents of failure.

The blown fuse was returned to the manufacturer for inspection. Based on the manufacturer's analysis is of the fuse, information was provided that the fuse opened under load, and that there was no apparent evidence of any defect within the fuse. Therefore it is presumed the fuse performed as designed. The manufacturer concluded the fuse was progressively weakened by repeated breaker closures until it opened to clear the circuit.

• Adverse Condition Report 92-277 LER NO: 261/92413, 014, 017, 018 B-57 PRELIMINARY

PRELIMINARY Work request WR/JO 92-ALHYI has been initiated to inspect the breaker to determine if any function of the closing operation of the breaker could have caused a condition of excess current draw sufficient to blow the 10 amire fuse, and to perform any necessary maintenance to correct such a condition.

The fuse manufacturer has recommended to use a LPN-RK fuse in DB-50 breaker closing circuits. This recommendation has been entered into the H. B. Robinson Technical Manual/Vendor Recommendation program under tracking number 92-0140 where it will be appropriately evaluated trough the Fuse Control Program as a possible alternate fuse selection.

V. ADDITIONAL INFORMATION A. Component Failures None B. Previous Similar Events None LER NO: 261/92-013, 014, 017, 018 B-58 PRELIMINARY

PRELIMINARY LICENSEE EVENT REPORT (LER)

FACILITY NAME: H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO[DCKET NO: 261 TITLE:

UNUSUAL EVENT DUE TO LOSS OF OFF-SITE POWER AND REACTOR TRIP EVENT DATE: 08/22/92 LER #: 92-017-00 REPORT DATE: 09/21/92 OTHER FACILITIES INVOLVED:

DOCKET NO: 05000 OPERATING MODE: N POWER LEVEL: 100 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR SECTION:

50.73(a)(2)(i) 50.73(a)(2)(v)

LICENSEE CONTACT FOR THIS LER:

R. D. CROOK, SR. SPECIALIST REGULATORY COMPLIANCE TELEPHONE: (803) 383-1179 COMPONENT FAILURE DESCRIPTION:

CAUSE:

SYSTEM:

COMPONENT:

MANUFACTURER:

REPORTABLE NPRDS:

SUPPLEMENTAL REPORT EXPECTED: NO ABSTRACT:

On Saturday, August 22, 1992, H. B. Robinson Unit No. 2 was operating at one hundred percent power.

At 1007 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.831635e-4 months <br /> a loss of offsite power occurred due to a trip of the Startup Transformer. The loss of the Startup Transformer caused a loss of Emergency Bus E-2 and Instrument Bus 4, causing a turbine runback. At 1009 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.839245e-4 months <br />, a high level in "A" Steam Generator caused a turbine trip and a subsequent reactor trip.

At 1010 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.84305e-4 months <br /> the Auxiliary Transformer tried to transfer its load to the Startup Transformer as designed, and a loss of E-1 resulted. At 1012 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.85066e-4 months <br /> the Emergency Operating Procedures network was entered and immediate actions were begun for response to the reactor trip. In accordance with the Emergency Plan, an Unusual Event was declared at 1025 hours0.0119 days <br />0.285 hours <br />0.00169 weeks <br />3.900125e-4 months <br /> due to loss of offsite power.

The plant was stabilized and repairs were initiated on the startup Transformer.

The Startup Transformer trip vas caused by a short circuit in the sudden pressure fault protective relay sensing circuitry. -During the event, the plant response performed as expected. There was no threat to public safety since both Emergency Diesel Generators started as required and provided power to the Emergency Busses. Repairs to the Startup Transformer were completed and normal power vas restored to the Emergency Busses at 0050 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> on Sunday, August 23, 1992. The Unusual Event was terminated at 0124 hours0.00144 days <br />0.0344 hours <br />2.050265e-4 weeks <br />4.7182e-5 months <br />.

This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(C) and 10 CFR 50.73(a)(2)(iv).

LER NO: 261/92413, 014, 017, 018 B-59 PRELIMINARY

PRELIMINARY I. DESCRIPTION OF EVENT On Saturday, Augut 22, 1992, H. B. Robinson Unit No. 2' was operating at one hundred percent power, with no major evolutions or activities in progress. At 1007 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.831635e-4 months <br /> a loss of offsite power occurred due to a trip of the Startup Transformer.' The loss of the Startup Transformer caused a loss of Emergency Bus E-2 and Instrument Bus 4, causing a turbine runback. Due to the loss of E-2, Emergency Diesel Generator "B" started and loaded properly. The primary plant transient caused the Reactor Coolant System (RCS) inventory to shrink, lowering the level in the Pressurizer to below ten percent. At 1009 hours0.0117 days <br />0.28 hours <br />0.00167 weeks <br />3.839245e-4 months <br />, a high level in "A" Steam Generator caused a turbine trip and a subsequent reactor trip. At 1010 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.84305e-4 months <br /> the Auxiliary Transformer tried to transfer its load to the Startup Transformer as designed, and a loss of E-1 resulted, causing the "A" Emergency Diesel Generator to start and load as required. At 1012 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.85066e-4 months <br /> the Emergency Operating Procedures network was entered and immediate actions were begun for response to the reactor trip. A manual safety injection was initiated at 1018 hours0.0118 days <br />0.283 hours <br />0.00168 weeks <br />3.87349e-4 months <br /> due to the decrease in Pressurizer level and the inability to maintain level with the Charging Pumps.

Pressurizer level recovered within a short period of time and the safety injection was reset at 1021 hours0.0118 days <br />0.284 hours <br />0.00169 weeks <br />3.884905e-4 months <br />. In accordance with the Emergency Plan, an Unusual Event was declared at 1025 hours0.0119 days <br />0.285 hours <br />0.00169 weeks <br />3.900125e-4 months <br /> due to loss of offsite power.

As a precautionary measure due to the nature of the event, the onsite Technical Support Center and Operations Support Center were activated to support plant response. At 1037 hours0.012 days <br />0.288 hours <br />0.00171 weeks <br />3.945785e-4 months <br />, the safety injection was terminated. At 1052 hours0.0122 days <br />0.292 hours <br />0.00174 weeks <br />4.00286e-4 months <br />, the backup Pressurizer Heaters were energized from the emergency buses, and at 1103 hours0.0128 days <br />0.306 hours <br />0.00182 weeks <br />4.196915e-4 months <br /> Natural Circulation was verified with RCS temperatures stable at approximately 500 degrees F. The plant was stabilized and repairs were initiated on the Startup Transformer. At 1348 hours0.0156 days <br />0.374 hours <br />0.00223 weeks <br />5.12914e-4 months <br />, a deviation from Emergency Operating Procedure EPP-021 was taken in order to restore power to the Deepwell Pumps to supply the Condensate Storage Tank.

The NRC was notified of this event via the ENS pursuant to 10 CFR 50.72(a)(1)(i) as a declaration of one of the Emergency Classes specified in the licensee's approved Emergency Plan. The NRC was notified via the ENS of the procedure deviation mentioned above pursuant to 10 CFR 50.72(b()(i) as a deviation from the plant's Technical Specifications pursuant to 50.54(x).

II. CAUSE OF EVENT The start-up transformer trip was caused by a short circuit in the sudden pressure fault protective relay sensing circuitry. This short circuit was the result of water collecting in the base of the cable connector at the relay (see attached photograph). A cable connects the relay to a junction box approximately two and one half feet away, and about six inches above the relay. The cable houses three conductors which H. B. Robinson Steam Electric Plant, Unit No. 2, is a Westinghouse Pressurized Water Reactor in commercial operation since March, 1971.

2Adverse Condition Report ACR 92-307 LER NO: 261/92-013, 014, 017, 018 B-60 PRELIMINARY

PRELIMINARY connect the relay to the transformer protective circuitry. This cable is hollow with the conductors loose inside. The junction box, which is designed with a drain hole for removal of moisture, had been inadvertently rotated to the point where the drain hole allowed water to collect inside. The water subsequently entered the hollow cable and traveled to the base of the relay/cable connector, where it shorted across two soldered connections.

The reactor trip was caused by a high steam generator level resulting from loss of instrument busses powered from the start-up transformer.

III. ANALYSIS OF EVENT During this event, there was no threat to public safety since both Emergency Diesel Generators started as required and provided power to the Emergency Buses. In addition, the Dedicated Shutdown Diesel Generator was available throughout the event to supply power if called upon. Appropriate provisions are available in the Emergency Operating Procedures to control the Plant for an extended period of time until some form of AC power is restored (i.e., offsite power, Emergency Diesels, or the Dedicated Shutdown Diesel).

This report is submitted pursuant to 10 CFR 50.73(a)(2)(iXC) and 10 CFR 50.73(a)(2)(iv).

IV. CORRECTIVE ACTIONS Repairs to the start-up transformer were completed and normal power was restored to the emergency busses at 0050 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> on Sunday, August 23, 1992. The Unusual Event was terminated at 0124 hours0.00144 days <br />0.0344 hours <br />2.050265e-4 weeks <br />4.7182e-5 months <br />.

V.

ADDITIONAL INFORMATION A. Failed Component Information None B. Previous Similar Events LER46-05 Figure "Sudden Pressure Fault Protective Relay Circuitry (Correct Position)' omitted.

LER NO: 261/92-013, 014, 017, 018 B-61 PRELIMINARY

PRELIMINARY LICENSEE EVENT REPORT (LER)

FACILITY NAME' H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 22BDCKET NO:

TITLE:

DEGRADED CONDITION:

LOSS OF BOTH SAFETY INJECTION PUMPS DUE TO FOREIGN MATERIAL INTRUSION EVENT DATE: 08/24/92 LER #: 92-018-00 REPORT DATE: 09/22/92 OTHER FACILITIES INVOLVED:

DOCKET NO: 05000 OPERATING MODE: N POWER LEVEL: 000 THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR SECTION:

50.73(a)(2Xi)

LICENSEE CONTACT FOR THIS LER:

DAVID CROOK, SENIOR SPECIALIST REGULATORY COMPLIANCE TELEPHONE: (803) 383-1179 COMPONENT FAILURE DESCRIPTION:

CAUSE:

SYSTEM:

COMPONENT:

MANUFACTURER:

REPORTABLE NPRDS:

SUPPLEMENTAL REPORT EXPECTED: NO ABSTRACT:

On August 24, 1992, H. B. Robinson Unit No. 2 was in hot shutdown condition and preparing for startup. At 1826 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.94793e-4 months <br /> during performance of a surveillance test, the licensee declared Safety Injection pump "B" inoperable due to inadequate recirculation flow. At 2258 hours0.0261 days <br />0.627 hours <br />0.00373 weeks <br />8.59169e-4 months <br />, Safety Injection pump "A" was declared inoperable due to an observed declining trend in the pump's recirculation flow. With both Safety Injection pumps inoperable, Technical Specification 3.0 was entered, which requires that the plant be placed in cold shutdown condition within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The plant achieved cold shutdown condition at 0020 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> on August 25, 1992.

The cause of the Safety Injection pump "B" reduced recirculation flow is attributed to foreign material blockage within the associated minimum flow recirculation line flow orifice. This material had been previously identified and reported in LER 92-013. A system recovery plan was initiated, which included extensive system inspection, cleaning, and pump testing, and installation of permanent recirculation line strainers.

This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(A) as the completion of a plant shutdown required by the plant's Technical Specifications.

LER NO: 261/92-013, 014, 017, 018 B-62 PRELIMINARY

PRELIMINARY I. DESCRIPTION.OF EVENT On August 24, 1992, H. B. Robinson Unit No. 21 was in hot shutdown condition and preparing for startup following a reactor trip.' At 1826 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.94793e-4 months <br />, following performance of an unscheduled surveillance test to redemonstrate Safety Injection system operability, the licensee declared Safety Injection pump "B" inoperable due to inadequate recirculation flow. At 2258 hours0.0261 days <br />0.627 hours <br />0.00373 weeks <br />8.59169e-4 months <br />, Safety Injection pump "A" was declared inoperable due to an observed declining trend in the pump's recirculation flow.

Although the recirculation flow acceptance criteria was satisfied, after consultation with the licensee's Operations Manager, the pump was conservatively declared inoperable based on a greater than ten percent decline in flow rate from the last three tests.

With both Safety Injection pumps inoperable, Technical Specification 3.0 was entered, which requires that the plant be placed in cold shutdown condition within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. A shutdown was initiated and the plant achieved cold shutdown condition at 0020 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> on August 25, 1992. The NRC was notified of this shutdown via the ENS as required by 10 CFR 50.72(bXl)(iXA).

II. CAUSE OF EVENT Event investigation? has 'been completed.

The cause of the Safety Injection pump "B" reduced recirculation flow is attributed to foreign material blockage within the associated minimum flow recirculation flow orifice. Through tracing materials used on site, the likely source of the material and its system entry point were determined.

It was confirmed through interviews that during Refueling Outage 14, the construction crew on Modification 1087, RHR Minimum Flow Recirculation Line Modification, had experienced problems resulting from inadequate purge during the welding process. They employed the use of a plastic sheet material to attempt a mechanical line block, or purge dam. Four circular pieces were cut for use as purge dams to support installation of check valves RHR-782 and RHR-783. All of the pieces were taken into the RHR Heat Exchanger room, but only two were taken up the scaffolding to the immediate work area.

The line was sufficiently large to attempt the installation of these plastic dams, and they were taped in place inside the ten inch piping for RHR Train *A'. However, it was determined to be too difficult to obtain a satisfactory seal in the line with the material, and this effort was subsequently abandoned.

During completion of the job the material was used to protect the seats of the check valves during grinding work.

'H. B. Robinson Steam Electric Plant Unit No. 2, is a Pressurized Water Reactor in commercial operation since March, 1971.

2 Licensee Event Report LER 92-017.

3 Adverse Condition Reports ACR 92-249 & ACR 92-250 LER NO: 261/92413, 014, 017, 018

"-63 PRELIMINARY

PRELIMINARY It is suspected that pieces entered the RHR system piping due to breakage. Although the exact amount and mechanism of material introduction is unknown, it is suspected that a maximum of two discs (approximately 155square inches) may have entered the piping. Follow-up interviews and investigations were unsuccessful in quantifying the amount of material that entered the piping or the mechanism for entry. During closure of the line, Quality Control personnel employed the use of a camera to inspect the line for cleanliness. This was performed by inserting a camera into the vertical line, and looking down and up through the open check valve.

This did not include inserting the camera beyond the elbow below the valve, and they were not able to see around the elbow into the horizontal run. As such, the QC inspection did not detect the presence of any foreign material.

The modification was completed and the system refilled for testing and return to service. Acceptance testing for Modification 1087 operated the RHR system at various flowrates using various flowpaths.

During testing and operation, it is assumed that the material was pumped through the RHR system. It is further theorized that some of the material was deposited behind the SI-863A valve, which was a "dead leg" projecting at a right angle away from the main flow path during recirculation. This made a natural trap for the material. Later, when the cavity was drained, this valve was opened, and the material was swept toward the RWST and SI pump suction header. When the RWST level reached forty percent, cavity draining was suspended, and SI pump full flow was conducted. Cavity draining was then resumed.

The material was discovered during testing in July in the SI Pump "B" recirculation orifice.'

The blockage identified in August was thought not to be a new piece, but a residual that was too large to enter the recirculation line during July. It is speculated that subsequent use of the SI pumps eroded the material sufficiently to allow it to enter the recirculation line during August. It had been originally thought that the material was broken into very small pieces from the SI pump and the material would easily enter the piping. This observation was determined by the fragments found in the orifice in July.

No other material has since been recovered from the any of the SI pumps or associated piping.

The only other material located has been in the RWST as expected and previously communicated.

III. ANALYSIS OF EVENT le blockage of the limiting flow orifice in the Safety Injection pump recirculation piping prevented the minimum recirculation flows needed to assure reliability of the pump during periods when the pump is not flowing water to the Reactor Coolant System.

During periods of operation under minimum recirculation flow conditions, this recirculation flow provides the only source of cooling to the pump.

4 LER 92-013, Plant Shutdown Due to Safety Injection Pump Inoperability, July 27, 1992.

LER NO: 261/92-013, 014, 017, 018 B-64 PRELIMINARY

PRELIMINARY Evaluation of the chemical composition and physical properties of the foreign material found determined that, had the material entered the Reactor Coolant System (RCS), it would decompose.

No material remnants have been found, and there has been no evidence seen through sampling of a substantial deposition in the RCS.

This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(A) as the completion of a plant shutdown required by the plant's Technical Specifications.

IV. CORRECTIVE ACTIONS Adverse Condition Report (ACR)92-333 was initiated to document the unsuccessful efforts to remove debris from the Safety Injection system as initially identified in July, 1992 and documented by ACR 92-249.

Two teams were established for system recovery which was initiated in August, 1992. One team was established to determine operability and cleanliness of the Safety Injection pumps. The second team was to investigate the source, potential locations, effects, and significance of the foreign material. A single project manager was established for the total effort. Special procedures were developed to control work, responsibilities, and evaluation of items found. The reactor was to remain in cold shutdown until all activities were completed to ensure the reliability and operability of the SI System.

The recovery efforts were intended to accomplish the following:

o Identification of the foreign material.

o Identification of possible entry points of the foreign material, its possible present locations, and a method to retrieve or flush material from the system, as appropriate.

o Evaluate potential damage and assure potentially effected Emergency Core Cooling System (ECCS) equipment is operable and can be relied upon during any flow condition.

o Assure that the potential presence of foreign material will not impact the operability of plant systems or components in the fiure.

o Identify the root cause of the problem and the corrective actions which will be taken to preclude recurrence.

In order to faciitite identification of the foreign material and the potential impact it may have had on plant safety systems, visual inspections of the interior of tanks, components, and piping determined through evaluation to potentially contain foreign material were conducted.

Documentation of the evaluation of areas, piping, and components determined not to require visual inspection was also prepared. These areas included:

LER NO: 261/92-013, 014, 017, 018 B-65 PRELIMINARY

PRELIMINARY o

The Reactor Coolant System o

Portions of the Residual Heat Removal (RHR) System o

The Chemical and Volume Control System Purification o

The Spent Fuel Pool Cooling System o

The Charging Pump Suction o

Portions of the Safety Injection System o

The Containment Spray Pump Eductor The components inspected included:

o The Refueling Water Storage Tank, (Using Divers and Cameras) o Both SI pump Minimum Flow Recirculation Line o

The SI Pump "B" Discharge o

The SI and Containment Spray Pump Suction Line o

The Spray Additive Tank Flow Transmitter o

Piping From the RWST to the SI-862A Valve o

Containment Spray Pump Discharge Lines As a result of the RWST inspection, cleaning of the tank was performed. For Safety Injection Pump "B" the piping and orifice were removed and the source of blockage was determined to be one thin piece of white plastic, approximately one-half inch in diameter, identical to the foreign material discovered during investigations in July 1992. Analysis of material confirmed it to be Derin, the same material found in previous investigations.

Plant Modification M-1134 was developed and implemented to install permanent strainers in SI pump recirculation lines.

Original plant design did not provide equipment to prevent plugging of the recirculation line flow orifices. These strainers, which would include flush and vent valves for each SI pump recirculation line, would serve to facilitate removal of any foreign material that should enter the system, and prevent the orifices from plugging.

A high velocity flush of each SI pump was conducted to provide assurance that the pumps were free of additional foreign material. The SI Pump vendor was consulted, and full flow testing of each pump was conducted on August 30, 1992 to assure no damage effecting pump performance had occurred as a result of the passage of the material through the pumps, or as a result of running the SI pump "B" with inadequate recirculation flow.

The inspections discussed above showed that the Deirin material was only in the RWST and SI pump "B. Since none of the material was found in the SI pump "A', the decision made regarding the trend seen during the previous flow tests was considered to be conservative with respect to the condition of the SI pump "B". Evaluations and tests of choke points and system interconnections reveal no other places where Delrin, if present, could cause a significant safety problem.

Pump and valve tests have demonstrated acceptable performance of equipment, and cleaning and flushing of piping and components has assured that the material should not reenter systems or components.

LER NO: 261/92-013, 014, 017, 018 B-66 PRELIMINARY

PRELIMINARY All results, evaluations, and conclusions were reviewed on September 10, 1992 by the Plant Nuclear Safety Committee-priorto plant restart.

V.

ADDMONAL INFORMATION A. Component Failures None B.

Previous Similar Events LER 92-013 LER NO: 261/92413, 014, 017, 018 n-67 P RELMINA RY