ML19350A151
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{{#Wiki_filter:- UNITED STATES , p'8 % NUCLEAR REGULATORY COMMISSION q ADVISORY COMWrTEE ON REACTOR SAFEGUARDS /,s n,k-()gh n O / I f r f CASHINGToN. o. C. 20555 Oe,g. %@> [f February 20, 1980 7 , J/2p j ' ll The Honorable Morris K. Udall, Chairman Comittee on Interior and Insular Affairs House of Eepresentatives Washington, D. C. 20515
Dear Congressman Udall:
27, 1979, you expressed the hope that the study of In a letter dated July Event Reports by the Advisory Comittee on Reactor Safeguards would address the consistency of actual component failure experience (e.g. Licensee You also asked the valve failure rates) with that projected in MSH-1400. ACRS to determine the probabilities of occurrence that, prior to the events, would have been predicted for the sequences of events that occurred 24, 1977 and at Rancho Seco on March 20, 1978 at Davis-Besse on September In a letter dated on the basis of ESH-1400 failure rates and methodology.the ACRS advised you August 15, 1979, detailed response to your requests and that it hoped to be able to complete ( this effort in approximately six months. in the calculation of the probability of an event sequence, retrospect, is ill-defined, since it depends entirely tpon the ensemble of Of course, W is letter event sequences in which the one under discussion is embedded. includes what are thought to be reasonable judgments on this point, and the results depend upon these judgnents. With the aid of the NRC Staff, the ACRS invited a large number of institu-tions in the U.S. and abroad, including the Electric Power Research Insti-tute and the U.S. reactor vendors, to provide data and analyses responsive including the NRC Staff itself, have Several groups, to your request.sutnitted component failure rate data developed since the compilation was We NRC Staff have sianma-made for the Reactor Safety Study, MSH-1400. rized the new data in Table 1, which also provides the failure rates usedSome of th in MSH-1400 for the same components and systems.in Table 1 is plotted . Also able spread in data obtained and the relative position of M SH which is astrated in Figure 2. Only plants which reported any failures are shown in Figure 2; hence, some plants had much higher failure rates 3D/ Ego 7f
y . February 20, 1980 khe Honorable Morris K. Udall than MSH-1400 on certain components while other plants had no failures Although to some degree the observed during the reporting period studied. a certain variation may reflect actual differences from plant to plant, portion of the variation may be due to differences in the repo to differences in the responses of reporting personnel. Turbine-driven purnps generally exhibit a higher failure rate (a factor of 10 h e NRC Staff is now giving extra attention to 100) than used in ESH-1400.Furthermore, a large variation in diesel reliability to this specific item. was observed among the various plants. the uncertainties in failure rate data are The NRC Staff believe that in MSH-1400, and that the general trend is larger than were projected heir preliminary assessment is toward somewhat higher failura rates. that this might produce an increase in their best estimate of core melt probability by about a factor of three. None of the groups who were invited have provided probabilistic analyses, using MSH-1400 failure rates and methodology, of the Rancho Seco and 20, 1978 and September 24, 1977 respec-Davis-Besse transients of Marchtherefore, asked three ACRS Fellows to devote effort We ACRS, tively. comensurate with the time available to provide such analyses; the results ( of their study are included as Attachment A to this letter. It is The ACRS believes that the results they obtained are reasonable. clear that the manner of treatment of human error can have a very large Also, for the Rancho Seco transient, the effect on the results obtained. aumerical results are very sensitive to the context in which failure of control system power is calculated. We ACRS Fellows also estimated a probability per reactor year of occur-rence of the major sequences which were present in the %ree Mile Island 2 accident of March 28, 1979. Of some interest in this regard is an observa-tion by representatives of Electricite de France that by applying M SH-1400 methodology they would calculate an overall probability of the order of for 'IMI-2, but when the events were connected by strategic opera-3x10 -7 tor errors, they found a probability as high as 6x10 -3 had several institutions provided independent We ACRS anticipates that, estimates of the probability of the two transients, a considerable varia-tion in their answers would have been likely.
F =- L. . February 20, 1980 We Honorable Morris K. Udall ~ Although the NRC Staff did not analyze the probability of the Ra provide the ACRS with two related memoranda, which are enclosed as Attach-ments B and C for your possible interest. The ACRS trusts that this letter is responsive to your request. Sincerely, Milton S. Plesset Chairman ACRS Fellows Report, " Analysis of Feedwater Transient Sequences in B&W Attachments: A. Nuclear Steam Supply Systems," February 7,1980 Nuclear Regulatory Commission Staff Report, " Evaluation of Davis-Besse and Rancho Seco Feedwater Transients on 9/24M7 and 3/20M8 Using B. WASH-1400 Data" Memorandum from F. Rowsome to R. Fraley, "ACRS Query on Material Davis-Besse and Rancho Seco Transients," C. Relevant to Udall Letter: (' February 12, 1980
TABLE l 'J
SUMMARY
OF CURRENT ^.URE RATE DATA SURVEY m ISD (2) 439 443 45) te) (73 (Cp (37,3 ARn. LER tout us GENE RAL tVALuATION ifEEJ.TEL SISLIS AfDMIC PROGRAM NCSR tePRDS WASH-1400 WESTINGHOUSE VOLTA OAnklEK. CtNtPONE NT F AIL s00ht 8.7E-5 2 0E-5 +1E-2433 1 2E-5 AUX FEkD PUMPS Fall TO START SE-443/80) F AIL To hun ECCE PUMPS FAIL TO START es.st-3 +3E-343/109 42 7E-3 +20-5 3 9E-5 +3E-383) +tE-3 4 E-3(33 42.3-129E-d 3. 5E - 4 3E-Stt09 te-200lE-6 3E-Sil0) FAIL TO Ru" 1 5E-5A F AIL TO ST ART & RUN IE-4433 +1E-3438 4.3E-e +1E-4438 ftANuAL VALVES 'F AIL TO OPERATE F AIL TO MMAIN OPEN (PLUG 8 (2 5E-6A } t it-50DE-4 4tE-343) {+1C-343B} 42E-3 tit-53E-3 48E-343) ft09'O FAIL TO OM M (SE-* 43-8003E-7 3 5E-et103 FAIL TO CLOSE SPURIOUS OPERATI0st 4.5E-5 4tE-343) SE-e ALL NGDES tlE-343) +5E-4 etE-3(3) SOLENGID UALVES FAIL TO OMN tSE-4 3.5E-et3D FAIL TO CLOSE 41E-3438 3.5E-7 (2-63E-e SPURIOUS OPERATION 9E-6 FAIL TO OPERATE +t2-5003E-5 +1.2E-Stas +1.5E-3 t ( 2-50 )E -3 +3.2E-542) GI>FLult VALVES F AIL TO OPEN FAIL TO CLosE +1.3E-2 5.5E-et2) IWURIOUS OPERATION 2E-563) +3(-4438 ALL le0 DES 2 2E-6 43&-543) +7E-3 3E-St3D WACuun VAL'JES F AIL 10 OPEN +1E-4tlOlH +44-7)E-3 1.4E-6 ftE-St3) SE-e +5E-4 +1.3E-3(10D S.4E-e el.0E-2410) RELIEF WALVES FAIL TO OPEN + 3E-2 8 3 )H 443-4)E-3 1.2E-e F AIL TO CLOSE SE-4 10E LIGHT 3E-4 3.5E-et3D IE-5638 101 HEAVT RE-St3)H 3E-6 PMMATURE OPERATION 4.5E-6
- IE-2 PI'4T MLIEF ULVS FAIL TO OPERATE FAIL TO RESEAT 7E-4 SELLOWS ML VLV FAIL TO OPERATE flE-Sf3) 2.1E-7 flE-4837
+ SAF E T Y VALVE FAIL TO OPLM 3.5E-et3D FAIL TO CLOSE SPURIOUS OPERATION 2.9E-et1.2) CHECK VALVFS (<EVERSE LEAKAGE +5E-5 1.4E-5 +1E-443) 45E-3 +1E-443) 3.ef-7 3E-+ ifiOD SE-7 3.et s FAIL TO OPEN +2E-4 8.4E-4 SE-5 FAIL TO CLOSE et.2E-5 SPuRIDHS OPERATION +1E-4139 3.2E-5 FAIL TO OPERATE RE-194309/E 1.6E-10/S SE-SS g 2 4E-9/8 1.5E-5 1E-tot 303/fi PIPES >3* RUPTURE 3E-104100/303/8g ALL NOSES 3E-9430) PIPES <3* RUPiuM 3E-104800/308/S, 2.4E-9/S 7.7E-e 1E-94303/S SE-7 IE-9 i.eE-10/8 ALL plu M S y 1 9E-7 +1E-4(3) til-25DE-4 ele-443D SCRAM RODS FAIL TO SCRAN 1 9E 5 +3E-443) ELECT. CLUTCH F AIL TO OPLRATE SE-ello) PHEMAluhE DISEteSAGEMENT 4E-6 +3E-4438 ptECH CLuf tti FAIL 10 OPERATE PFtJMATURE DISENGA0EftENT 4.2E4e ,,GAShLTS t.E Ah AGE J.SE-4 kup DAIVE FUNCTION Sheet 1 of 2 CONikot
'.I A o
- til (2)
(3) (5) tel 475 E09 tCl ll0f PICCARSe LER L0tfE Ases SISLIS ATOMIC PROGRAM NCOR MPRDS WASH-1400 UESTINOHOUSt VOLTA GARRICE DEC*tTEL SEMERAL EVALUATION EssentesENT FAIL seOSE 1.eE-et' Sl 7E-e 3 5E-7 t*'E*8t'*88 f SATTERY ALL teOSES f' 480 OUTPUT 42E-e SATTERY SYSTEM FAILURE 881 SEftANS 3E*et3D FAILURE et-et 3/3 9 /F 7E-4 1 ALL seOSES 1 5C-e t 2,.9 9 8; 2 2E-4 + 3E-24 3 9 3E-5 +45-50DE-3 +3E-2439 f9.N-3D 8 SATTERY CHARSER ALL ft0M S 3E-38109 3 3E-4 7E-4 3C-3tIO) 1 3C-58 CIEM L KasERATOR FAIL TO START +4.2E-3 +3E-242/103 +2 2E-2 +3E-2 JE-4439 f FAIL 10 RUBf 3,3g-3 DVERALL FAILURE f +tt-800lE-4 +2 3E-4tS.9) 5.SE-7 +5E-e el.0E-ett03 CIRCUIT SREAktRS F AIL TO OPEN I.2E-4 SE-6 4 3C-St109 F AIL TO CLOSE IE-et3D +4E-4fS.99 SPURIM OPERATION RE-et3D 3.2E-7 2 3E-e 7.5E-7 +1E-3439 +3C-4 FAIL TO OPERATE elf-343)
- g. 7.E -7 3 3E-7 2 7E-e (2-53E-4
+3.5E-4443 +1E-4439 ftELAYS FAIL TO OPERATE 3 4E-7 43-10lE-7 5.M-Sf 353 FAIL TO ENER0lZE SPURIOUS OPERATIO88 * +1 5E-St71 1 i feAlt, SWITCHES FAIL TO OPEN {+ tE-54 3 D} +5.0E-9te7.1 i L e.4E-0(l.4i FAIL 10 CLOSE SPURIOUS OPERAT8086 3E-7 FAIL TO OPERATE +1E-4(3) TSRSUE SWITCHES Fall TO OPERATE 2E-743.39 3 5E-5 4tE-443) 9.4E-St34.5 PRESSURE SeftTCH F AIL TO OPERATE i PMMATURE OPERATIcel +2.tE-est.9) to.2E-7tt.9) LIIIT SWITCM S FAIL TO OPEN 4.2E-641.99 FAIL TO CLOSE +3E-443) +2E-5 EPuRIOUS OPERAT1000 2 5E-e FAILURE TO OPERATE (4E-e 45-10 L-4 4 4E-442 7T 3 5E-5 L33. EEV. MisSOR FAIL TO OPERATE PMMATURE OPERATIOst <eE-7 35-2 Sit-4 1 7E-745.7) 3 4E-5 PftESS. SEssSOR F AIL 10 OPERATE OUT OF LIMITS gg.58909 45-109t.-4 1 5C-et5 0) 7.5E.5 TE8EP. SENSOR FAILURE Out Or LIMITS 3g.5439 a. s j
- 3. A "+" preceeding a failure rate denotes failure-per-da NOTES.
- s. LETTER Surrans Oss rR LURE RATES Mn0TE THE r0Lt0Ulwes A - UPPER 951 CONFISEteCE SOUMO All other failure rates are failure-per-hour.
C SAnERT CHAR 0ER 8 - Z E g,9 S - FER f>ECTION OF PIPE S - FOR SIZE CLASS 1750-2900 f(W SIESEL-SEMERAT08.3 H - F AILURE DAT A FOR HELIUM 2 THE PLUMBER OR FIUMSERS 388 PARENTHEMS Fot.LOUINS FAILURE RATESME48e9 DNE SHOULS MMOTL THE Rase 0E FACTORS. FOR EXAMPLE (NP/TT pMu 10 OSTAIN THL UF FER 952 Mut TirtY THE HEDIAM VALUE ST
- eEDIAN St VV TO OleTAIM THE LOndLR THE CONS 1pfMCE DOUND ANS isIVII4A SIfe0LE NUMSER IN PAREdelHESES INDICATES St E ONf I DE NCE DOUNie.
UPPEft ANie LUteER le0theb. FfE RANGE F ACTam IS FOR SulH THf Sheet 2 of 2
DEFINITION OF TERE Biblis Nuclear Plant in Federal Republic of Germany ( Biblis - Provided by the Natiorcl Center of Systems Reliability - United Kingde NCSR Provided by Dr. Guiseppe Volta - Ispra Volta Provided by Licensee Event Reprt Data Evaluation LER Provided by General Atomic Company GA Pickard-Provided by Pickard, [ owe, and Garrick Motor Operated Valves MOVs Relief Valves RVs Diesel Generators DGs Circuit Breakers Gs Iagend: WASH-1400 O ~1 10 Volta Volta LERs 5 _2 Biblis g 10 Volta g f4 D. 1 -3 g 10 () g Biblis 0 0
- s Pickard O
O C -4 10 Volta. Volta GA ,NCSR -5 10 -6 10 Volta i i i l Pumps MOVs RVs DGs Gs Scram Rods i Fall to Fail to Fail to Fall to Fail to Fall to Start Open Open Start Close Insert Figure 1. Data Point Estimate Extremes ) 1 l ?
'r Imgend i b - - H ER hM2 ) O wnSa-1400 Ranga DEFINITION OF TERMS -0 10 HDVs - Motor Operated Valves ~ M. Pumps - Motor Driven Pumps T T. Pumps - Turbine Driven Pumps l DGs - Diesel Generators I I I l I i I I -1 10 i i v I i I I I I I e i
- C) e I
I I 4 m I I I l 1 1 -2 I 10 I 2 I 1 E i l O ( I l l l E l 10 1 () () O ~ 'A> ~4 10 i i PC/s M. Pumps T. Pumps DGs IXhs Fall to Fail to Fail to Fail to Fall to Open Start Start Start Run (. Figure 2. Plant to Plant Variation ~~
.,?
- o..
ACRS Fellows Report, " Analysis of Yeedwater Transient Sequences in B&W Nuclear Steam Supply Systems," February 7,1980 ( i, Attachment A . - _ _}}