ML19332F485

Review of Use of PRA in Integrated Plant Safety Assessment Study for Oyster Creek Reactor, Final Rept Phase I

ML19332F485
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Site:	Oyster Creek
Issue date:	12/31/1984
From:	Davis P NEW JERSEY, STATE OF
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{{#Wiki_filter:,. f l. ? FINAL REPORT PHASE I I A REVIEW OF THE USE OF PRA IN THE INTEGRATED PLANT SAFETY ASSESSMENT STUDY FOR THE OYSTER CREEK REACTOR l "t Prepared for [ THE NEW JERSEY BUREAU OF RADIATION PROTECTION t l. t [ By t. P. R. DAVIS P. O. Box 1604 Idaho Falls, ID 83403-1604 l 4 8912150068 891211 December 1964 j

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-p W ~ g ?' CONTENTS Section Page I-INTRODUCTION 1 11

GENERAL COMMENT

S 2 111 SPECIFIC COMMENTS 5 IV CONSIDERATION OF SEP TOPICS WITH POTENTIAL 11 FOR RELEASE FROM LESS SEVERE ACCIDENTS V CONCLUSIONS 15 i REFERENCES 16 p 6 1 P i b

ik, p ~ g ' A REVIEW OF THE USE OF PRA IN THE INTEGRATED PLP/4T' SAFETY ASSESSMENT. STUDY FOR THE OYSTER CRELX REACTOR I. INTRODUCTION. This' report presentsL theyresults of a selective review of the NRC document-entitled " Integrated Plant Safety Assessment - Systematic Evaluation Program,'- '0yster Creek Nuclear Generating Station" (NUREG-822)U) The review focused

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~ p, lprimarily.on-the validity of the use of probabilistic risk assessment-(PRA) in the NUREG document. Specifically, the main objectives of.the review were to: Assess the validity with which PRA was used to establish i risk significance of the issues considered; Evaluate use of PRA to derive backfitting decisions with [: respect to plant. safety. j p i; 7 [. A secondary objective.was to: u

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' Identify issues which.would have an impact'on.the.proba-I bility of offsite release less significant th'n considered , in PRAs but-in excess of protective action guides, q

The.NRC study for Oyste.r Creek is part of t. larger program undertaken by the

= NRC, called !the Systematic' Evaluation Program, in which an NRC review is.under- / taken of t. der nuclear plants to reconfirm and document their safety. The review provides (1) un assessment of t1e significance of, difierences between - g current NRC technical' positions on saf)ty issuesand those positions -that existed when the Oyster Creek plant wa:. licensed (0yster Creek was granted ~ an.. operating license in 1969), (2) a basis for deci. ding on how these differ-ences should be' resolved-in an-integrated plant review, and (3) a documented eviluation of plant ' safety. I .In= undertaking the SEP,'the NRC developed a list of 800 topics which were

used to evaluate each of the older plants considered in the program - Somo Jof ~ these topics -includeo several related issues which are evaluatad sepa-rately. ;In the Oyster Creek study, the NRC determined that of the 800, only 40 were determined to be candidates for in-depth evaluation in the SEP. The; remaining 760 were eliminated based on various evaluations, 1

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. y, m i{f f ~ i -including dup.lications,' applying only to PWRs, considered as part of other NRC programs-for evaluating' plant safety, a determination that.0yster Creek was in compliance', or some other basis. m The scope of-the review covered in this report includes only an evaluation of _ the finalf40 topicsland their-related issues '(a total of 87 issues). Thus,. .a. ~this report'will consider the use of PRA in evaluating;these 40; topics, except~ ~ f that--some comments.of--a-more general nature are included as derived from the

review.

l 1'! -The format of the report' includes a section of general comments (Section.Il following), a section (Section Ill) of specific comments for-each of the 40 l topics which were evaluated with PRA, a discussion of those topics 'which Lhave the potential for influencing the probability of release in excess of protective: action guidelines (Section IV). and a final section (V) of con-clusions. An evaluation of the risk signi1icance of the 40 topics which were not evaluated with PRA in the SEP was not-considered in this-review butzis ' considered in a subsequent report (13) r Til.- GENERAL COPNENTS ~ 1. - The original Phase 1 topic list which. includes 800 items was reduced - y*>

to 13'7 for consideration in the Oyster Creek SEP. The basis for elimina-L tion of the 663 topics is not provided in the SEP except for some gen-I eral elimination criteria on Page 2-1.

While there is no reason to i s ? suspect that any of the 663 topics eliminated should have been re-- l ,tained because of important safety implications at Oyster Creek, a j w 1 listing of'them'and the basis for their individual elimination is not m i explicitly provided. e Of the remaining 137, 24 were deleted on the basis that they are being f E considered under other NRC programs (per Appendix B). This appears to ) y? 'be a reasonable basis for elimination, but it should be recognized that some of these items remain unresolved with respect to their implication-on Oyster Creek-safety. [ k 14 g .n,

Of the remaining 113, 30 were deleted on the basis that they "did not apply to Oyster Creek" (pg. 2-4). This evaluation is provided in Appendix C. In reviewing the Appendix C list, it appears that the 30 deletions are appropriate. tiowever, it is not true that none of the ),. 30 apply to Oyster Creek (as stated in Table 2.1, pg. 2-4). Five of the 30, according to Appendix C, were deleted on some other basis, including being resolved under other generic activities. Of the remaining 83, 43 were deleted on the basis that they "... met current criteria or were acceptable on another defined basis" accord-ing tu Table 2.1. This table refers to Section 3.1 for the deletion of these topics. However, Section 3.1 does not provide any further basis for the deletion of the 43 topics, and no individual considera-tion of them is provided. In many cases, it is not clear why the topic was deleted. 2. Of the 40 topics retained for evaluation in the SEP for Oyster Creek, 20 were examined (in whole or in part) on the basis of risk signifi-cance using PRA methods. These 20 ind ved a total of 27 issues, 21 of which were PRA evaluated (Appendix D). In five cases, the topics included one or more issues which were not evaluated with PRA. Of the 20 topics, 17 were found to have a low impact on risk, two a medium impact, and one a high impact. However, of the 17 found to have a low impact,13 were identified as requiring "backfit", which includes further study and evaluation. The three topics rated ( " medium" or "high" were found to require backfit. Of the four topics determined not to require backfit, none were eliminated solely on the basis of the " low" risk significance determination. [ [. Other considerations were also a part of the decision to eliminate l these topics. Thus, in the Oyster Creek SEP, it can be stated that none of the SEP topics were deleted only on the basis of PRA evalua-tion and that of the several hundred SEP topics deleted for Oyster Creek, only four were deleted partially on the basis of explicit Oyster Creek risk significance considerations. L a

,w 2 ~r I l 13. .Several deficiencies were found in evaluating the PRA study provided in the SEP report (contained in Appendix D). These include:

a..

Recent component and system _ failure data were generally not. . used in the PRA aspessments., Instead, reliance was placed. on the WASH-1400(21 data base which is now over 10 years old. Much effort has been expended (most of which has-been spon - sored by the NRC) to update the WASW-1400 data (for example, see.Refs.3_through 7). b.- -.It appears that common cause failure contributions have been ignored in some of the evaluations. c. The basis for assumptions and various values used is fre- . quently not provided, d. The significance of simplifying assumptions is frequently - not stated, and no uncertainty assessment is provided for any of the evaluations. The existence and impact of these deficienciet are considered when -app'ropriat'e for'each of the SEP topics evaluated. ' n view of the very minor role which PRA played in determining the ~ 4. i -disposition of the SEP topics (see comment 2 preceding), it appears -that the Appendix D_ assessment is reasonably appropriate as a screen- .ing evaluation. Furthet, the-deficiencies in comment 3 above do not appear significant'except as noted in specific cases covered in the; next'section. 5. 'The basic approach in Appendix D was to examine each of the systems -(components, procedures, etc) affected by the issue and determine if it appeared in a dominant accident sequence of the Millstone Point I IO) a plant similar to Oyster Creek. If it did not risk assessment ( appear in' a dominant sequence, it was dismissed as " low" significance. The possibility that the issue could increase the failure probability of a system in a non-dominant sequence and thereby cause it to become significant was apparently not evaluated. It does not appear that this apparent deficiency would be significant, however, because most system failure probability changes are minor compared to the change required ,to cause a non-dominant sequence to become significant. A further ~ g m

g j v ] ' evaluation of this issue will be provided in a subsequent part of' the 1 review in which it is planned to examine the Millstone point I _ risk ~ pc.

assessment study with respect to its-implication on SEP topics for Oyster.

-Creek.. 111.; SPECIFIC' COMMENTS This section provides an assessment of each of the 21 issues which were evalu-ated by theLPRA assessment described in Appendix D of the SEP report. p ' 1. Table Ex-2 of Appendix D lists the. 20 topics and associated issues and thbir risk classification (high, medium, or low). However. in Table 4.1, five of the 20 issues considered have not been given a "PRA Rating" (last column of Table 4.1). These include topic numbers IV-7. A.3, -VIII-4;. - XV-16, XV-18, and XV-19 (all rated " low" in Table Ex-2). p, 2. The evaluation of each of the 20 SEP topics on the basis of risk -significance _in Appendix D was reviewed. ~The results for each topic are as followsi i Topic 111-8.A; Loose parts Monitoring and Core Barrel Vibration a. Monitoring - This assessment appears valid for loose parts moni-toring. There appears -to be no evidence that a ' loose parts moni-toring system would influence overal1 risk. It: should be.noted g that the Appendix D evaluation applies only to loose-parts noni-O toring and not core barrel vibration monitoring, even though: L (pg. V, Appendix D) it is' concluded that the entire issue has i been classified'as low importance to risk. However, it does not appear conceivable. that a core barral vibration monitor. would have any influence on risk since core barrel-vibration-induced I severe accidents have never occurred. b. Topic -111-10. A; Thermal Overload protection for Motors of Motor-Operated Valves - This. assessment utilized WASH-1400 data for motor-operated valve failure rates (1E-3/d) and Nonelectronic parts Re- ~ liability Data to assess the contribution from thermal overload-protection. failures (bypass of thermal overload protection dur-ing accident conditions is now required to eliminate this failure p~, ~ d c m

mode). The conclusion from the evaluation was that only a small effect on risk woult: occur by imposing the bypass requirement. This assessment was checked by reliance on more recent valve failure data for nuclear plants as compiled in Reference 4. The Reference 4 derived failure rate for MOVs is 4E-3/d, some four times higher than used in the Appendix D assessment. Referente 4 also lists and quantifies the contribution from 28 failure mech-anisms. No contribution from failure in the thermal overload protection is specifically given, and a contribution of only 4% was found for electric motor operator failure (these failures pre-sumably include thermal overload protection failures). If thermal overload protection failure problems were important, they should have been evidenced in the Reference 4 survey since most of the 8 data are from non-accident conditions wherein thermal overload protection is not bypassed. It is thus concluded that the Ap-pendix D conclusion for this topic is valid, c. Topic IV-2; Reactivity Control System including functional Design and Protection Against Failures - The Appendix D conclusion appears valid. .d. T_o_pic V-5; Reactor Coolant Pressure Boundary Leakage Detection - Appendix D concludes that this requirement would have a low impact or. public health risk. Part of the conclusion is based on the assertion that PRAs have not shown small break LOCAs to be risk significant for BWRs. To check this, three BWR PRAs were exam-ined; Browns ferry (9), Millstone 1(8), and Peach Bottom 11(2) In no instance in these studies were any LOCAs determined to be risk significant (except those induced by transients). Fu rther-more, no PRA has givan credit to the possibility that a leak detection systcm could reduce LOCA probability. Thus, augmenting the Dyster. Creek system would appear not to result in a measur-able risk reduction. The Appendix D conclusion appears valid, e. Topic V-lD.B; RHR Reliability - The Appendix D conclusion appears valid. ( - a

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Topic'V-ll.A; Req $irementsforIsolationof.High-and_ Low-Pressure Systems Several problems were found.with the Ap-pendix D. assessment of this' topic, as'follows

_m ~ ')._ The assumption'of'a core melt resulting from RWCU. iso 16 tion - failure is not realistic. Several tore cooling systems ~ would. remain available'following the event. 11. The. Table'l' data from WASH-1400 do'es not agree with more recent data:. s Table I Reference 4 Check 3E-7/hr. 7E-7/hr -Valve-Leakage' ' Relief lE-4/d 8E-3/d(BWRs) s n - Valve-L . Fails to 'j b ~ Open-he 'iii~.-._ The Figures 2:and 3 fault trees; appear to contain errors. For-example, the failure rate for " Pressure-'. Sensor IJ04 fails"should be a demand ~ failure and not a per yr. failure.- -since-it-is only' involved.a f ter an initiating event -(valve - failures). It is not clear. why, failure values -are different l for -"CV does not-reseat" on Fig. 3. Also.f the' pressure sen- ~ ~ sor failure value on Fig. 3 is different~than Fig..2.- However, in view of"the fact-that a core melt'.would not ensue- ~ b for this event (RWCU failure) the Appendix D conclusion appears.- f valid. Further,ithe NRC has, required the_ applicant to perform an analysis'of the system (pg. 4-27,228) and make system modifi-r cations as necess uy. It is concluded by the NRC on Page 4-281 ^

that the resolution of this issue is of highlimportance'to risk h

because of the possibility that (1) the. relief. valve may not have-sufficient' capacity and (2) the potential-for pressure sensor '~ failure leading to a large 'LOCA. Neither of these' aspects were evaluated in Appendly 3 It appears that these considerations [ are valid,and the _fina. NRC position is considered appropriate. U 4

^ + v tk ' Whise' the potential to increase risk significantly does not appear ' high fron RWCU failure, the probability of low level offsite re- [ 1 ease may be a concern (see Section IV). It is recommended that i the resolution of this topic be further examined as the applicant's -study becomes available (according to pg. 4-28, the licensee has 1 agreed to submit the results of. the study to the NRC staff by January 19831). i 1 i g. _T_opics VI-_4; Containment Isolation System and VIII-4; Electrical ~l penetrations of Reactor Containment - The Appendix D assessment [ concludes that reducing containment isolation failure and penetra-P tion leakage would not impact risk since these leakage modes have i [ not been found-to be important risk contributors based on several BWR PRAs'. This assessment appears correct. .i p 1 h. Topic VI-7.A.3;.ECCS Actuation System - The Appendix D evaluation o of-this issue is somewhat confusing. It is concluded that emergency I y condenser testing during operation is possible without taking both I ( units out of service. However, n evaluation.is.provided_on whether such testing should be requireJ e" what risk reduction could accrue .as a result. To help resolve this topic, the dominant accident sequences from the Millstone I study (8) were examined. It was-found E that emergency condenser failures are relatively significant con.. tributors to risk.: Thus, a reduction in emergency condenser failure could reduce risk. However,-it is not expected that increased test-e ing during operation would provide a significant reduction in failure probability since considerable testing is already accomplished ac-s cording to Section B.6, Vol. Ill- (pg. B.6-3) of Reference 8. Fu r-thermore, one of the dominant failure modes-(pg. 6-11, Vol. 1 N Ref. 8) for the system is being left in the manual (test) mode after testing. The system was found to be highly reliable under I o b current test conditions and requirements (pg. 6-11). 'l s 1. Topic VI-7.C.1; Appendix K--Electrical Instrumentation and Control o .(El&C) Re-Reviews - This topic was found to have a medium impact on risk as a result of the Appendix D assessment. However, numerous l ~8' E _

j-l} 4 45 r g deficiencies were found in the Appendix D evaluation (pg. 60, et. I seq.),~asfollows: i.;: No basis is provided for the short-to-ground failure rate of 3E-7/hr.

11. Breaker. failure rate is a sumed to_be 1E-3/ demand, but

( more recent evaluations (5 show a rate of'4E-3/ demand, k, iii. No basis is given for the assumption that the automatic bus-transfers have never been tested. iv. Tour. simultaneous breaker failures are given identical independent failure rates, but no contribution for com- .F mon cause failures is provided. 1 v. The basis for.a single diesel failure to' start and run (0.06) is not given, alth ent with recent data (0.7)ough it-is reasonably consist - vi..The assessment' (pg. 64) of two emergency diesel gener-ator failures includes only the-independent failure of -each~ unit (0.06 x 0.06 = 0.0036) with no ommon cause assessment. However, a recent evaluationt6) gives a failure rate for a 1: out of 2 diesel generator system as 0.001 to 0.007, which encompasses the value in Ap-pendix 0. . Based on the foregoing, the Appendix D evaluation of this topic appears flawed.- However, the NRC has. decided to require' the licensee study to establish corrective actions necessary to precl0de auto-transfer of faults, and such a study appears warranted (pg. 4-36). It is not 'coccted that this issue would be-risk significant since neither the Browns ferryI9)'nor the Millstone PointiI I ) PRAs found loss lof emergency AC power to'be significant. In the Millstone Point PRA, automatic bus transfers were assumed to exist at the plant. i j Topic VI-10.A; Testing of Reactor Trip System and Engineered Safety Features, Including Response-Time Testing - The Appendix D assessment appears valid. k. .Togic VII_-l_.A; Isolation of Reactor F mtection System (P,PS) from Non-Safety Systems - The Appendix D asse sment appears valid. p p

a 1. Topic VII-1.B; Trip Uncertainty and Setpoint Analysis - The Appendix D assessment appears valid. m. Topic VII-2; ESF System Control and Logic Design - The Appendix D assessment appears valid, n. Topic Vil-3; Systems Required for Safe Shutdown (f.lectrical) - The Appendix 0 assessment appears valid, o. Topic _ Vill-2; Onsite Emergency power Systems-Diesel Generator - The Appendix D assessment appears valid. An independent check of diesel generator failure causes from Reference 11 confirmed the Appendix D conclusion that protective trips are not major causes of diesel failures. p. Topic Vill-3; DC power System Bus Voltage Monitoring and Annunci-ation - The Appendix 0 assessment appears valid. This was the only topic assessed to produce a "high" impact on risk. DC power reliability has been detennined to be important in other generic studies and is one issue being evaluated by NRC's Task Action plan (for Unresolved Safety Issues) A-44, Station Blackout (10) q. Topics XV-16; Radiological Consequences of Small Lines Carrying Reactor Coolant Outside Containment, XV-18; Radiological Conse-quences of Main Steam Line f ailure Outside Containment, and XV-19; Radiological Consequences of Loss of Coolant Accident - The Appen-dix D conclusion that these events do not lead to core melt is correct, According to Section 4 of the SEp report, the licensee has agreed to implement a CWR Standard Technical Specification Limit for primary coolant activity which should further limit re-leases for the first two topics, for the third topic, the licensee has agreed to develop and implement a maintenance program for main steam isolation valves or justify existing maintenance. e .m 4 ~ s y;. _ e kf'- {- # y. Qf'. b

IV.

CONSIDERA110N OF SEP TOPICS WITH POTENTIAL FOR RELEASE FROM LESS SEVERE ! ACCIDENTS-n d . f This section presents the results of a brief survey of the.40_ topics considered in:-the SEP: study; to determine _ if any have the potential for' increasing the-prob - [pf Lability of events leading to the release of radioactivity from accidents less severe than core melt' accidents. While;these events may not be risk significant in a public health = sense, as discussed-below, they may produce doses at or near - lthe site-. boundary sufficient to trigger protective action,- and they are therefore f of some interest. t<> Probabilistic.riskiassessment studies have consistently concluded that public -health'Eisk-(in terms'of early1f atalities, radiation injuries, and latent cancer ,4

deaths)-from ' nuclear power plant accidents are dominated by core melt accidents.

M F In~ other words, it has" been determined that the: product of. probability and con - I. sequences forFaccidents less severe than core melt do not; impose _ significant-L f public-health ; riskscom' pared to-the same product for core melt accidents. As' b ,a' resulttof this~ determination, PRAs dismiss non-co' e melt accidents, andLthey ~ r .were noticonsidered in the NRC's risk assessment determination for.the SEP top:cs, nor are they considered in' the Millstone Unit 1 PRA, or. other BWR PRAs. - b

Nevertheless, federr.1 and state agent.ies, including:the New Jerssy Bureau of--

J< Radiation Protection, are bound by= specific groundrules which require' protective-action!:to be taken when doses-at the site; boundary exceed certain levels LTypi-- cally,Jthese levels are very low compared to the-threshold for. significant health ' effects,.and,they: can'be exceeded-by accidentsLless severe than core melt events. %A 'etermination'of;the amount of radioactivity release required to exceed pro-- d

tsci.ive action guide (PAG) doses, and the probability.of such releases, was beyond s

_ :the' scope of'this study. - It was therefore not possible'to evaluate the 40 SEP topics?with respect to-their relative potential for causing events resulting in

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ddses exceeding PAG limits.~

Instead, the topics were screened to determine S

which appeared to have a direct potential for influencing the probability of L

1 E T events' producing' doses exceeding PAG limits. - 1t "shoul'd be recognized that = the,NRC's _40 SEP topics evaluated for Oyster Creek ~ did noti consider all' sources of radioactivity within the plant which could, if ~ ~ a

u. b_' c released, cause the PAG doses to be exceeded. In-a typical BWR, some seven y distinct sources of radioactivity esist, and they all vary greatly in intensity 7 .' depending on plant activities and the particular time.in the fuel cycle. These sources are the reactor core, the reactor coolant, shipping casks, refueling transfer casks, waste gas storage, and liquid waste storage. The 40 SEP topics ' ' considered releases from only the first two (risk assessments consider only the core). Thus, for this screening study, only the reactor coolant radionuclide release potential will be considered since core melt releases are covered under c e her c re d ccur u c e el a e.

In order for primary coolant or partial core damage releases to exceed site

' boundary PAG doses, it appears necessary that the release must occur outside the containment-boundary. This is because the plant is designed to meet.NRC . site boundary dose' limits for the worst accident involving discharge of the entire reactor coolant inventory inside the containment plus a substantial radionuclide release from the-core. This is shown by results in the Oyster Creek FSAR-(Ref.12) wherein site boundary doses (1/4 mi.) of 0.18 rem whole ~ a body.and 0.049 rem thyroid are calculated for worst weather assumptions. These L calculations also assumed very large core release fractions (100% of the noble-l gases, 50%.haolgens, 50% volatile solids). These. doses are far less than the lower limit New Jersey PAGs of.1 rem whole body and 5 rem thyroid. - The-40 SEP: topics were, therefore, screened to determine which could have a direct- -influence on-the potential for the release of the reactor coolant. inventory and

core damage release outside the containment boundary. The NRC evaluation of backfit action'for' these topics was also reviewed to determine what reouirements, if any,Lwere being imposed.

e The most credible way for a radionuclide pathway to the environment to develop J is :for a breach of the. reactor-coolant system to occur outside containment. If this occurs, radionuclides in the' reactor coolant, plus any others added as a result of core l damage, can be released to the environment. Thus, the 40 SEP ~ 3 topics were screened to d2termine which have the potential for contributing to'such an event. It should be emphasized that a release of coolant outside containment does not necessarily mean that PAG limits will be exceeded. Site bou.ndary doses under these conditions depend on many fact ., including amount r

and type of radionuclide inventory in the coolant at the time of release, rate and total omount of coolant release, weather conditions at time of release, and location of release. An examination of these factors, including estimates of site boundary doses, is beyond the scope of this study. The following topics were found to have the potential for release of reactor coolant outside the containment boundary. A brief discussion of each is also included: 1. Togic 111-5.B; Pipe Break Outside Containment (Issue #2, Emergency Condenser Isolation) - This topic and related issue considers the possibility of a rupture and isolation failure of the isolation condenser steam line, producing a release of primary coolant out-side the containment. 2. Topic V-5; Reactor Coolant Pressure Boundary (RCPB) Leakage Detec-tion (Issue #3, Intersystem Leakage) - This topic relates primarily to leakage inside the containment. However, the third associated issue (intersystem leakage) appears to provide the potential for leakage outside the containment through interfacing systems. 3. Topic V-ll.A; Requirements for Isolation of High-and low-Pressure Systems - This topic considers the fact that Oyster Creek does not comply with NRC requirements for independent interlocks on the reactor water cleanup system which transfers reactor coolant water outside the containn.ent. 4. Topic XV-16; Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment - This topic relates to the NRC requirement that the radiological consequences of failure of small lines carrying reactor coolant outside the containment should be limited to small fractions of the exposure guidelines of 10CFR100. According to NRC calculations as described on page 4-45 of the SEp (Ref.1), the potential offsite doses "would substantially exceed the applicable dose limits". !li

f. 5. -lopic XV-18; Radiological Consequences of a Main Steam Line O Failure Outside Containment - The NRC determined (pg. 4-46, p L Ref.1) that the radiological consequences of a steam line break outside containment do not meet NRC criteria of " limited to small fractions of the exposure guidelines of 10CFR100". 6.- Topic XV-l_9; Loss of Coolant Accidents Resulting From Spectrum ) of Postulated Piping Breaks Within the Reactor Coolant Pressure h Boundary - The NRC staff calculated (pg. 4-46, Ref.1) doses in excess of 10CFR100 guidelines for the thyroid for-the design 7, l-basis loss-of-coolant accident. The major fraction of the dose F (334 rems of a total of 341) was due-to MSIV leakage. (These results appear to be far in excess of the results in the Oyster Creek FSAR described previously for the same assumed accident. The reasons for the difference are unknown, but appear to be related to assumptions regarding MSIV leakage.) For 'all of the above topics except Topic V-5 (#2), the NRC has required and the utility has apparently coninitted to some backfit activity. Of particular interest -in this regard is the intent by the utility to reduce the Technical Specification- -limit for-reactor coolant activity. The extent of_ this reduction-and the result-

ing influence on dose reductions is unknown.

However, the reduction will have an effect on all of the topics listed, and the'NRC has determined,'according to the SEP report, that such a reduction will-satisfy the requirements of Topics XV-16.18, and 19. It is not known, on the other hand,_ whether such action will result in doses below the New Jersey PAG limits which are lower than 10CFR100,- During the review of tile topics to determine _ potential _ influence on low icvel

• releases, two_ additional topics related to containment-integrity were identified which could influence low level release probabilities.

Loss of containment in-tegrity. will not cause any significant release under normal operating conditions, .but a loss of this integrity at the time that a radionuclide release occurs-inside containment (from a LOCA, for example) could have a very significant bear- -ing on do3es. The two topics are: '+ Topic VI Containment isolation System Topic Vill Electrical Penetrations of Reactor Containment. "I4'

n ,1 i p r The NRC has required some minor backfitting for Topic VI-4, but not for VIII-4. ItLis of interest to note relative to containment integrity that, according to I page F-xiii, Appendix F of the SEP report (Ref.1), an increased frequency of containment-integrity violations has been noted in recent years at Oyster Creek, V. CONCLUSIONS p. L Based on the review as described _in preceding sections of this reporti the fol-lowing-conclusions are drawn: a

1. -

for a large number of deleted SEP topics, the NRC evaluation (I) does not_ provide the basis on which they were deleted for the Oyster Creek plant. .2. The quantitative probabilistic risk evaluation of SEP topics was a very minor consideration in deriving backfitting requirements. [N ~ Instead, _it appears that the-primary basis for decisions on-backfit requirements was engineering judgment coupled with the extensive experience base developed by the NRC in using the design basis ac- - cident concept in conjunction with deterministic analysis, .f -3. The risk assessment evaluation of 20 SEP topics as provided in Appendix D contains several apparent discrepancies, in no case, = however, were such discrepancies found to'have at adverse influence on final decisions made regarding backfit requirements. The ~quali-tative results in Appendix D used for backfit decisions-are, there-fore, considered valid- / 4.. Several topics were _ identified which appear to have an influence on- . the probability of. causing releases below levels considered in PRAs, but in-excess of Protective Action Guides. + P b ') p - .- e j

Mm'e: yn. 9: -4 1 5 7 REFERENCES 1.

1_ntegrated Plant Safety Assessment - Systemtic Evaluation Program, Oyster Creek Nuclear Generating Station, NUREG-0822, Final-Report,

Ja nuary.1983.

y 2.

Reactor Safety Study, WASH-1400, U.S. Nuclear Regulatory Commission, October 1975.

j 3. Data'Sunmaries of Licensee Event Reports of Control Rods'and Drive' Mechanisms at (T.'STTommercial Nuclear Power _ Plants. NURET[CR-1331, T6&Tidaho, February 1sMi, ' flu _ clear Power P'1 ants, liOYEDKR7763~p_ orts of Valves at U.S. Commer Data Summaries of Licensee Event Re 4. , E'GRIldaho, June IT80. 5. National Reliability Evaluation Program (NREP) Procedures Guide, RURTG7CR-2B15. l6.. " Reliability of the' Emergency AC Power System at Nuclear Plants",

R.; E. _8attle, et.al., presented ~at International Meeting on. Thermal -

Nuc1carReactorSafety, Chicago,IL,1982(NUREGCP-0027.). 7. Data' Summaries of Licensee Event Reports of Diese1' Generators at U.S. ConimercfaTTuHear~ Power PlantfNUREb7CTf-TJ6~2, EG&G 1daho, March W80.'

8. -

- Interim Reliability Evaluation Program:. Analysis ofLthe Millstone. ) ^ Point Unit 1 Nuclear Power Plant",f Science Applications, Inc., Feb; ruary'1983. 9 ~. Interim Reliability Evaluation Prog _ ram: -Analysis of the Browns-Ferry, i 70iiTtTNuclear P1ent, ~ NURTG7&T802TITMTdaho, July 1T82.

10.'-

Unresolved Safety Issues Summary, NUREG-0606, Vol. 6, No, l', february 17, ,1984. 11.- Enhancement of On-Site Emergency Diesel Generator Reliability, NUREG/ fMT660.. University of DaffoHesearch Insti6ite, February 1979.. '12. - Oyster Creek Nuclear Power Plant, Unit 1, Final Facility Description. - and-Safety Analysis Report, Amendment 3, Part 1, Vol.1, January 1967. 11 3.. An Assessment of the Risk Significance of SEP issues for Oyster-Creek (Phase 11), P. R. Davis, December 1984. '. LS

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APPENDIX-A u

. 3 Kw f ..ANiESTIMATE:0F. RISK-FROM CORE MELT ACCIDENTS. E -:AT THE-OYSTER CREEK REACTOR 1 I 4his Appendix presents the'.results of. an analysis to estimate the risk' of'early; h>' fatalities =and 1atenticancers.as thelresult of. core melt accidsnts at Oyster

i

-Creek. :It should be vmphasized that the. estimate is very. uncertain -ir +5at it c s_ t "] ' depends on many. reasonable but geneially unverified.asa:cmptions and theiextrapo-- { ilation 'of otherJinform: tion,.the accuracy of which has not-~been f ully, verified. iThus, the result' can:be-considered:no 'more than a scoping estimate sutsject to ~ llargeuncertainties. ] ' The approach usediin producing the estimate consisted of; the ;following steps: j \\; n:,: El;. LThe core melt probsbility'as a function of radionuclide release category Mas estimated for-0yster Creek-by. adjusting ine results of:the Millstone-- ' Point Unit 1 risk;studyU ) to account for known? design'diffarences which ..were: judged 'to:be significant. i 4 L2.~ LThetradionuclide release' fractions for each category of accidents as:used 1 Sin the Millstone essessment-were estimated by.u'se of information in the~ l Reactor Safety [St.udyf 2), as modified by more're' cent information on source terms (3) ~ 3' ~ ^ s p 13. ' The, consequences of-a ~ core melt-:i.n 'the release categories were estimated l iby using the.results of a study performed byLSandia Laboratories ;for the g - LNRC(4) Njustments were made in the Sandia results to account for the J0yster: Creek design p;rameters and release characteristics, i, j 4 4. ' By combining the results-of:1, 2, and. 3 preceding, risk to the population m ' surrounding-thel 0yster Creek site in terms of early fatalities and latent 4 cancers was estimated. Each of these. steps will be explained in -detail in the following sections. 4 's,

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yp+ ; 4 i l."- ESTIMATE 0F.0YSTER CREE! CORE MELT ~ PROBABILITY-PER RELEASE CATEGORY: ~ 1The Millstone Unit =1-riskLstudyh) estimates core melt probability as a-function-ioff release category;for tLe Millstone 1 plant. By examining the design features - Uw Lof Millstone I: compared to'0yster Creek, it was determined that important design t features of-the' plants were 'similar except' that Oyster Creek does not-have.a -

feedwater coolantlinjection (FWCI): system which can be used -in Millstone I to-t supplyJ coolant ' tol the core. :This -design comparison ef fort was not rigorous tor -

' = ' comprehensive, 'and a major unverified assumption in the. study was that 'nosother, l design l1 differences of significance exist.- ITh'e Millstone core melt probabilities were adjusted.to account for the lack of-

a FWCI. system at 0yster Creek. ' This-was done by_ examining each core melt acci-m,

~ Edent_ sequence (sunnarized.in Table 8.3 of the Millstone report)-and determining ~ dn'whichcases1the: FWCl _ failure had an'important contribution to reducing the ~ sequence probability. The FWCI-function occurs in a.large number of the 25 s dominant = sequences for Millstone. However, in' many cases, failure of the system was caused by preceding; failures of support systems (predominantly electric pcwer). ItWas eventually determined that only three accident sequences would be appre-: ciably affected by the absence of an FWCl; system at Oyrier Creek. The proba-bilities off these _three sequences were requantified ~by removing the FWCl failuie probabii.tyTmul tipl ier. A more-detailed discussion of this procedure may be -found in Reference S. Th'e contribution to each release category of these'se- 'quences was recalculated, and the. total core melt' probability fo'r each frelease category was summ d. The result is shown in Table A-1. 'able A-1 COMPARISON OF MILLSTONE I AND OYSTER CREEK CORE MELT PROBABILITY M "' FOR EACH RELEASE CATEGORY 4 a_,7 bf Plant

RETea~se dategory ProbabilitEper yr.

TotalCoreMeltProbability( ~ 3 4 i 2 ~ 2E-4 3E-4 ~ 'E'- 6 8E 4~~TE-4 TMillstone j 1 Qyster Creek, _lE-6 8E-6 1.2E-4 3.2E-4 4.4E-4 I i

The re, case categories-shown in f able A-1 refer to specific fractions of radio-1 F

fnuclides contained in the core which are predicted to be released to the atmosphere. L i ~

~ y g-33-, i =-+ n

These categories also

implyLcharacteristics of the release, such as' timing, b._

elevation,setc.; lThe mEth6d__of[ deriving release categories originated.in the 4

q, _ P{eac. tor" Safety Study (2) and was ' retained in the' Millstone ' assessment.: 11t.was y"'o ifound ~during-the?RSS effort that the large -number.of accident sequences-examined < Lcould be groupedj into a few categories based on: similerities with ret.pect to - ~ @~ estimated radionuclide release parameters..Thus, release _ categ'ory I signifies; pg a particularf set of release parameters, and the probability of such a release cis)theTsum of the:probabilityiof-all? accidents found which were estimated to= W M* m

produce 'a! category 41 " release -(in this case,1x10~6/yr for _both Millstone. '^nd a

[' OysteriCreek);- Details; of = the, release parameters for each category;are< dis- ~ cussed::in thei next section.t 4 fili l ESTIMATE OFLRADIONUCLIDE-RELEASE PARAMETERS FOR EACH RELEASE CATEGORY R $. [1n;the?ReactorSafetySt0dyI2), four relcue categories were1 determined to q f adequately represent"the radionuclide release parameters during : core melt wQ Jaccidentsl for the Peach Bottom:II' reactor, a: boiling water reactor' similarito- ~ M LO/sterLCreek.1The RSS made -the further implied assumpt' ion' that:these same releaselcategories were sufficient to' estimate radionuclide-release from all i O) BWRs.- LThey were' also used in the assessment of the Millstone. reactor ,.a

reactor.very simi'*.r to Oyster Creek (5).

lt has, therefore, been assumed that these same' release' categories are appropriate for the Oyster Creek a'ssessment-

>7 . presented;herein.-

V ^.

y 9"~ - (The RSS1 characteristics of each of the1four release parameters are shown in - h LTable m2,Ltaken f rom WASH-1400. The table also shows a-more recent' assessment n of; some of the same parameters for release categories II and' 111. (No estimatss ~ m were given in Reference 3 for Categories 11 and IV.)- 1Some of the-parameters in Table A-2 require explanation in order to understand yp? -the_subsequentiuse to be maae 4 the i1 formation. The time of release column provides an estimate of-the elapsed time from the initiation of the assumed accident until.'the~ release to the atmosphere commences. This time is essentially the time calculhted -forecontainment f ailure to occur following the accident 'ini-i ~tiationffbr accident sequences in each category. The third column is an estimate

,; r ef-4the(le.ngthLof oti!ne the release lasts and depends on the assumed mode of j

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-m -- -containment failure. The " Warning Time" column is the estimated time from failure of the core-cooling system (s) which are expected to lead to core melt / until the' release occurs. It is assumed in the RSS that the operators would immediately be aware of such failures and take prompt action to initiate evacu-ation. The release elevation is the height _ above ground level that the release is assumed to occur, and the' energy release is the amount of energy contained in the radioactive plume that is released per hour. The remaining columns provide the fraction of each radionuclide species con-tained in the core which is estimated to be released when containment failure occurs. These columns provide the release fractions estimated in the RSS as well as, in some cases, an updated assessment, BMI-2108(3) The BMI fractions were provided for only BWR release categories II and 111 and do not include It has been determined (2,4) that fractions for the Ba-Sr, Ru, or la groups. radiation-induced health effects-(consequences) from the postulated accidents are dominated by three radionuclides, iodine, cesium, and tellurium. The BMI results do provide release fractions for these important species. The BMI results also do not include release fractions for organic forms of iodine ~(ninth column). However, these fractions are so small compared to other iodine forms (tenth column) that their contribution is negligible. In estirnating the risks from Oyster Creek, as will be shown in subsequent sec-tions, the BMI fractions were used, when available. Otherwise, the RSS fractions were used. III. ESTIMATE OF CONSEQUENCES OF CORE Meli ACCIDEi.TS AT OYSTER CREEK Sandia National Laboratories has performed core melt accident consequence calcu-lations for all reactor sites in the U.S.N) In performing these calculations, the folic..ing assumptions were made: 1. The core radionuclide inventory was equivalent to a 1120 MWe PWR. 2. A " swr.ary" evacuation model was used (people with 10 miles of the site move at 10 mph af ter delays of 1 br (probability of 0.3), 3 brs (probability of 0.4) and 5 r.rs -(probability of 0.3)).

w. 3. Actual site population (1970 census) and wind rose data were used for each site. 4. Best estimate regional meteorology was used. 5. Assumptions regarding dose / response relationships. dispersion models, etc. are as modeled in the CRAC2 computer code (3,4), 6. The source terms (release fractions) consisted of three core melt re-lease categories designated SSTl, SST2, and SST3. These categories were previously defined by the NRC in Reference 5 and apply primarily to PWRs. Consequences were computed at each site for each of these release cate-

gories, i

7. The consequence calculations for each site (including Oyster Creek) were performed assuming a 1200 MWe reactor, and the health consequence indices were early fatalities (deaths within 1 year of exposure), early illness (non-cancer sickness appearing within 1 year of exposure which require-medical treat'nent). and latent cancer fatalities. The release categories used in the Sandia study are shown in Table _A-3. The same information is provided in Table A-3 as was described for Table A-2. The Sandia study consequence estimates for the Oyster Creek. site are shown in Table A-4. i As can be seen in comparing Table A-3 and A-2, the SST release characteristics are different from those judged applicable to Oyster Creek. Thus, the Table A-4 I conseauence calculations need to be adjusted to account for these differences. Further, thc Table A-4 results are for a core radionuclide inventory correspond-ing to a 1200 M e reactor atkr than the Oyster Creek power of 620 MWe. First, the power level difference will be considered. Figure A-1 shows a plot of the relationship between power level and consequences based on information j fron, the Sandia siting study calcu'ated by the CRAC2 code. These resul ts are i for assumed releases at the indian Point site, but on a relative basis they I A-6 j R_

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sygyw% L w ^ ~ ' ^' N y :i * ~ y_ g' .:. N S,.. F.4 > kN a n Table A-4: ?* m ^ CONSEQUENCE CALCUlAT10NSLFOR OYSTER CREEK FROM 'y SANDI A;SITlNG STUDY , - "y y xf vi ' x,. ~. $ ". w Mean Early-Latent. Cancer-g _ v: Release Category ~ Fatalities _ < Fatalities-R

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/SSI1_ 84: 34400 ~ SST24 '0 200:.

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SST3:

-0 0.6

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= Y pW' jshould b'e similar for the '0yster Creek 1ocation. The - power ;1 evel - es sentially.. e iest'ablishes i.he radionuclide inventory. in the core and,, therefore,La reduced: opowerdevel reduces the amount of each radionuclide species released (if 'all-s_ Jotheria s sumptions - rema in; u ncha nged)_. Tnerefore, a reduced power level -shouldi A i,,~ Chaveja?significant1effect on'. reducing: consequences. - This: relationship is-Tillus - e>v T Ltrated(tiy pigure-A-1.. g e {^ < : Based 'on Figure. A-1I, it is possible to.. derive a reduction ' factor for' estimate'd. 20yster' Creek iconsequences 'from an SST1J reledse :to apply to the Table-LA-4; results.- L hibfactor.i'si merely the: ratio 1of the 1120 MWe consequences (used te comput'e thet T Stable A-4-results) ~to the Oyster Creek power level. From FigurefA-1 these ratios ~ u ' are: O - g b Nrimean--latentfatdlities: f . % 5"f"qu n 7 sL

• 0.75

= g j.. 295 620 W e consequences li36; = 0.36 - -ror mean early fatalities: 1120 MWe consequences p o m .. he; rJe).ttad,iustment' considered to Table A-4 is. related to.the physical character-T g istics' of tne release '(the.2nc through Sth columns cf Tables A-2 and A-3). As 1 m scan tie'seen in c'omparing these tat' lea, some of the characteristics are not equivalent as shown in Table A-5. Thus, these differences need to be considered m i sinsorder tc' edjust the-calculated Opter Creek consequences, if appropriate. h k E. i [' Q:, .j u.- c.. N a

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ggme p y , y g. ^ l g gM ' ' .n-w y i ~, ~ -); f bSb h, cy, n fs Table A-5 '[' [C01 PAR'ISON- 0F PHYSICAL: CHARACTERISTICS OF. RELEASE',- 41 -BWR-RELEASE CATEGORIES VS SANDIA'RESULTS'- 3 '?,

/

1. ~'

- Phvsical_ Char ' Time of j Duration of,a_clenistics_: Warning' Elevation _= m Release : O -( ' ? Categories, Relcasei Releases Timei j Release H b .Hr ~ _H r-l + M

• Btu /Hr.

1 ri 4-M 1 2 2' 1,5-25__ 2130 4 s l BWR-- 2.L 30. -3 ~ 2 .0 '30 'E -3 30 3 2 -- 25-- -. Categories 4 5 -' 2 2-25 10 2 - ~ - - 0 ~ SSTl 115-2 0.5 W 'Sandia SST2 3 - 2 1 10 '0: Resul ts ;. SST3-1 4 0.5-10 0 x

The' time of' release as sh'own in Table.A-5 is con,siderably longer _forL the BWR qcategories (especially:2-E. 3) than 'for the -Sandia results.

Hewever, the time iofarelease:only' delays the.. release and reduces the! fraction available'because i; Lof 0 radioactive. decay..This is already accounted for since the BWR release-fractionsLare, according-to the RSS, fractions relea ed to the atmosphere. L u / The{ duration of-release:(fourth column) is.: essentially:-e_quivalent for; all re-c ? lease 1 categories. UThe warning time-is somewhat shorter for the Sandia releases. This paramete.r: influences the consequences in that a longer time available for evacuation:means-R

that more people could.be removed from the plume pathway.

However, it.is judged that the Sandia times are more realistic since the times used for the BWR release ~ categories,Las defined-in WASH-1400, ara times from the failure of_ the system d.hich leads' to ucore. melt'.. No time penalty is provided to account for delays in [' diagnosing the plant state or-delays in transmitting and evaluating the decis-ion -to evacuate. Thus, it was judged-the Sandia times may be more representative, ~ ialthough> possit41y -conservative,.'. shen applied to the Oyster Creek consequence 4; f = estimates. ? e i ^ h. ( i 'k if ' I

yw ;,: R ~ {$$ g}- 3 j

.j 2'

xThe elevation ~ of release for the BWR categories is 25 m for all

but-Category 2 W

f (0heters). The Sandia1results all assumed ai10 m release elevation. However, r 'a ~ all)oftthese elevations ~.are relatively low, and elevation effects' would not be f^

significant(10)-_. - Thus,. n_o, adjustments appear necessary for these differences.

~ ~ S LTheienergy of.the release (last column) is a measure of the thermal' energy-. con-4 gu-7 -tained in the _ plume _ as it exits the containment. The. plume energy-can be;impor.- d Ltant in that it influences buoyancy effects and thereby-controls the elevation [ offthelplume' \\.The Sandia siting: study (4) examined this-effect, and theirzresults [ ihave been plotted' in Figure A-2 where the mean latent' cancer fatalities and early l fatalities are plotted against the energy release rate for an SST1 release. (The: P LSSTli release i.s appropriate for this comparison since its release fractions are i much closer to the BWR categoriesLthan any other Sandia category.) 'f From Figure: A-_2', adjustment ifactors can be estimated f_or application to the' Sandia - g k

consequence calculations for 0yster Creek (Table A-4).to account' for the plume

energy release rates of 0, 20,- 30, and 130 Btu /hr as shown'in column. 6.

For_-the 0 energy release -ca.se, __no adjustments are needed since the Sandia results are for ai 0 release.

For the 20 Btu /hr case,- the mean early fatalities need to-- a be. reduced bi a factor of 1 N = 0.53, and the mean latent fatalities increased- ~ 22 _c 300 byia factor of j g = 1.08. Factors can be readily derived-in a similar manner q f for the 30-and 1 0 BLu/hr cases. These results are shown in. Table A26. o Taole A-6 ADJUSTMENTtFACTORS'TO ACCOUNT FOR PLUME ENERGY RELEASE P. ATE-FOR OYSTER CREEK CONSEQUENCES ? -.InefgTIEfeas~e7 Adjustme.n_t Factors Q1_ l ~ RateJBtu/hr) 'iarly f atalities l latent Fatalities i' 0! 1 1 hf=.0.53 hh=1.08 20 h = 1.09 i30 l 0.5 = 2 h = 1.13 1130; [ 0.41 = f(lh ;DeriVeTfNiingir~c~4-2 I b JA-11

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-.--..L b:-

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$W e ~ }&jy y p f M ^ fyn + 6LpA - 1 JTheffinal,(and~most significant,: adjustment l to the Sandia: consequence calcula-y!:9f his for Oyster Creek arises from a= difference in the radionuclide release k[ fractions. :The Sandia' release fractions are~ basically for assumed PWR acci-dent coaditions;;and they -do not match the BWR releases:which have been assumed l for this'. study.1These differences can be seen by comparing Tables A and' A-3, i

These tables show -that-the. Sandia SSTl release fractions are closer to all:- four m

cof: thelBWR release fractions than any'of:the other SST categc.-ies..Iherefore, 4 um m L, the' SSTl releas'e fractions were used as the basis to adjust the Oyster Creek Q fconsequence calculations. f( The. adjustmentsLwere performed by thel following procedure. First, a direct com-E* iparMon' between the SSTl and BWR' releases was obtained by constructing ratios of the two releases for each radionuclice species. "The 'results are shown is, y - - e LTable' A-7,. _ Next, a plot was constructed showing the relationship between the reductiun in radionuclide fractions released and fatalities (both early.andi ' late). This pl'otLis'shown:in Figure A-3 ard is based on data provifed in the- ~ I4)

-Sandiaistudy

. - he' data are :for an SSTl release, ar.d: the -fractional releases 1 assumed ' apply-to all ' radionuclide species except-the nobles-; gases -(Xe and' Kr), dTheinoble gasirsleaselfraction was. maintained constant at 1.0. V Finally, by: use:of Table A-7 in conjuocticn with Figure A-3, adjustment ' actors =were estimated to account. for the difference in SST1 and the RSS/BMI_ release-bi '

fractions.for BWRs.

In Table A-7, the fourth column-indicates that the ratio 4 'of RSS/BMI release fractions to those nf SSTl varies from 0.56 to l0. -However, L ' bothothe early and latent fatalities are dominated by I, Cs,.and Te:(see Ref. 2,. Apprneix VI, SectionL13). For these species, the ratio varies from 0.63 to 1.09. 1 -For early fatalities, I and Te are more important than Cs, so these ratios are 'more significant. Thus, a source term reduction factor of 0.9 was -selected as ] an adj: stment factor for Category 1, early fatalities. From Figure A-3, it ^ ' can be seen that an SSTl release fraction of 0.9 reduces the early fatalities : 'I lby a-f actor of 0.9. Thir factor was used in the calculation of Oyster Creek o gm *. risks for early fatal-ities from Cctegory I releases. =I ) /For latent. fatalities, Cs dominates (Ref. 2) followed by I and Te. Accordingly, a reduction factor of '0.~8 was selected. From Figure A-3, an SSTl f raction of 1 I.8 would mean a factor of about 0.96 applied to the latent fatalities. By a 0 U A i g

I

( J

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• - +

mc s ,jm 7.. n u. f-m y ~.y .. i - - V, * %;@$?j -, LCOMPARIS01 0F;RSS/BMIfAND!SSTl. RELEAS.E1 FRACTIONS 1 l . A *; ~. 4%e

• ~

,~ m,, . f. we v r 4 - ~ * , t + ' w; s m,, g4 - r~ m z r ~,g e y e - I I'. 7 -III.

I c

-ly : slp P -RSS/BMI M Radionuclide L R557; ~ _... c~tio. TfT1Ra.-RS578MI RSS/;- . R ..RSS/BMI

RSS/c Ratio' SS/BMIi RSS/

Ratto 'BMI.

SSil-lJBMI-SSTP "BMI Ratio;

' Y Y1 .j SSTl

BMI-SST1-Species Xe-Kr 1.0 1.0 ;.

Ela cl.0 ~l... pl l 0l , r-wc 1., 9 4 ?- 1 0.6) l0.6) .7 8 ~ 1 a w -c ~ 0.44L

0'.0008

'O.001'81 kN. I 0.'45 ' O.4 '0.89-L 0.2'5

0.56 10.2l.1-I 3..-

Cs-Rb '0.67 0 4- .0.631 O.25 DG.37:. 0.2t 10.301

0.0057 (0.0075) ;

k ) 0.55 0.2-. J 0.31,. 0:004.

s Te-Sb 0.64 0.7-1.091 0.35; 10.0063 l-Ba-Sr

-l 0.07; 0.05: 0.71 0.1 Ll.432 0.01

• 10.14f 10.0006

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1. ~

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.m "a - ' .i ll@tlW %, c t; if s g ~ ..a s, w 'b }.M' /.' -g ~ p;-;f(. s ,ww. ? .Y 1 - i isimilar procedtfrehfactors were derived to apply to. early~ and latent ~ fstalities a n- ~( 3orLCategories}ll, Lifl,;and IV'.; ;Tuble A-8'.shows-the zresult's of 'this1 procedure.- 2 u 6s 1 - ~ e -Table A-8 ED "2 'EARLY1AND' LATENT FATALITY < FACTORS T0 ACCOUNT FOR; C @s RELEASE-FRACTION. VARIATIONS l = ~ w ~, c . (p - -7 Release Categories- _] i] 4 1 It .: ] ] - 1;j . ly

SSTl

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0. 3 5 --

0.69-0.006-- LO.032E

w a-IV.. ESTIMATE 0F'SEVERI ACCIDENT R1SKS FROM OYSTER CRf.EX:

w 1 LBy combiningitheladjdstments' discussed in'the-preceding section:to:.the--Sandia-tconsequence1 calculations for Oyster Creek?(Table-A-4), itlis possible to estimate risks' from operation of the _ plant due to severe accidents. J All;of the: adjustments' ~ . can-be. summarized in' two equations,L as follows: y m x 4.; ~ ~ (' f.e. L (Ei.e)(F i.e)(P )~ ~ R-Ea C P 1 e: - s,e': 0

^i' igl i

-< tend? m. 4 r t ,j B; c, Ni 'Rj = -- C3,)Pf,) .) (Ej,y)(F4g)(P ) l a ts tQ isi p- .Where: p; .y 4-is.the' societal (I) risk of early fatalities from senre accidents 'R>- ' e: i at Oyster Creek, per year, 9 \\.g -N1))sbciEtaltiSLs;are def,ined as the total risk to the exposed population fol-o i 11oktino ahr:acc:ident. i " .m m.. mw ~ J

4, , f d 6;a ' %i 8

• :. g -:

_ H ~ ,y h[_. '_ g i 1, ~ s ..) -h 4 gR ' : < - 'T: ? - =C are the; number of. early' fatalitiesicalculated to occur:at the. S' Oyster Creek site by the: Sand _ia! study;for an SST1 release (see. ~

i Table A.4).

.y . Pp 'EEis:af factor which adjusts for:the reduced powertievel at 0yster ~ Creck! compared to the power assumed in_the~Sandia' calculations. .(seeFig.-Afi). 1 . Ej . lis a factor. which adjusts ~.early fatalaty consequences due '.-- Eto-the energy' release of ;the radioactive plume-forJ the -ith l release category (I -through IV); see : Table-- A-6.- y - /1 f $ tis a factor which adjusts ear _1y fatality consequences due=to; ~ ' li F the-radionuclide release fraction differences between SSTl and: 3 [ .the BWR categories (see Table A-8) for the ith release category.. i th

P is;the.probatiility of a ~ release in the i BWR release-category g

for Oyster Creek (see Table A-1). .are all defined the same as above except J ~R),1Py*),C3*3, E,), F$') they apply 'for latent fatalities. 4 - Using.the'two preceding: equations-for1 R) and :R, along with values for: each e ~ term as derived in this' andLprevious sections. fit is possible to-compute. ,s estimated risks lfrom severe: accidents at Oyster Creek. TablesVA-9 and A-101 s provide-such aicomputation. TThese tables list valuesL for each of the factors ~ + - derived forleach BWR release category. The _ R : and R columns -in' these> tables - y e j 'showt the' actual

• values computed, and the last column'provides aLpercentage' contributionLtolthe riskf from each release category.

'[ Table A 9 ESTIMATE.0F EARLY FATALITY RISKS FROM OYSTER CREEK s,e {R _ l % rintribution ]:PfjC f'i, e P e l l 4 v I, 10.361 84-0.41i 0.9 I lE-6 1.1E-5 { l.5 M -11' 0.36 84 0.5 0.46 8E-6 5.6E-5 7.9. j 7-j 111-j0.36,,84 0.53 ! 0.32 1.2E-4 6.2E-4 87 e % 'IV 1'0.36 r j] up 84 ' l' O.00027j 3.2E-4 2.6E-5 3.6 3 .._f -i. l .i 10TAL

7. l E-4/yr !

100. ] ^ ayRx - 1 i {..gE . v 1 s 4 L

w%*

w

, ;p ;t :- J ciq ~ ~ ~ 11 m f h.} ' ~ m g.' t i ' l As{showniin Table A.-9fthe. estimated total mean'. number of early fatalities: lexpectedtol.the.populationsurroundingtheOysteECreeknsitefromsevere he. faccidsntsfisf 7l13E24/yr. Most of Lthis risk (87%) comes:from release cate-- ~ ~ ' " g111.f it, gor ^ cs e> e 2Similar1y, from Table A-10g;the total, mean number of ' latent-fatalities ex-- rpected to the p.o'pulation surrounding: Oyster Creek from severe accidents is' p(' j 0.=36/yr. Again, Category Ill. is the dominant contributor (83%). / Table A-10 q ESTIMATE OF. LATENT FATALITY RISKS-FROM OYSTER CREEK ~ C E F~ P 'T~ % ContributTon ~P fjl s,,1 1,1 i,1 i-1 1- -l 0; 75 : .4400-1.13 0.96 lE-6 3.6E 11 . v, 1 II. 'O.75 4400 1.09 0.78 8E-6 2.2E 6 4 ~@ sille '0.75 .4400-- Ll.08 0.69-1.2E-4 '0.30-83 2 x '!Vi f0.75 4400 1 0,034; 3.2E-4 3.6E-2 10 'u w. = _.. - - -:In oeder to-provide fome-perspective for these' results and'to'obtain a basis - [ (to? judge their significanca, a-comparison'was-.made with other risks,- and ant a . individual risk computation was performed. The comparison withiother risks gg 'was based on data from Reference 9.- m lTo calcul' ate individual risks, it is necessary to determine the population: around the-0yster Creek site which would be exposed to the radioactive plume go

ma Ereleased from'the severe accidents considered.

Based on results from the Sahdia tudy(4)., i_t appears-that some significant exposure could exist out to g pM fab 6ut-20 miles..Thus, a 20-mile radius was assumed in calculating the exposed j ! population. -- According to the Sandia study, the average population density out-E Jto.l20 ' miles from the Oyster Creek site.is 139 persons /mi2 (based on the 1970 icensus). LThus, the t; al exposed population would be (139)(r)(20)2 = 174,673 m h jpersons. Using this population in conjunction with the Table 9 and 10 results q p . Dand:

Reference:

5 : data, : Tables A-111 and A-12 were produced. y\\ Mb 1

y% : m g / h W : Wu.e www w i*,c wq Aqw ,2 + nw n o, t-Q Pdp;&p ~. -s. / F x_ s a - L. , -n m;, ~ ; ,jh{ " y,.., D ' WW 4Y%~~ yx 'g m,; W y* 4 , 4. w.- r n +L L M:.. M, 2:J 1 1Tabl'e~ A-ll : p,'.,,4-Q " LCOMPARISONL 0E OYSTER: CREEK! EARLY ~ y.,.. e 8 W.Q p%; - s 7 q Risk,- per yr. x;, Source 4 1 Societal.- <l Individual Q I R' ,10yster! C eekMearly. _7.lE-4 J 4'. l E-9 ; e. d!S ' fatalitie:; ' u s W ~ ~ /A' ciden.t'al',l U.S.\\ . 81 '. 6.-

4.7E-4 pf c

LAll;Caeses,1U.S.e

1534, 18.78E-3 83 ' ^

s M .c a Wi' y .Tabl el A-12 x ,0 ); ' F " s COMPARISON 0F10YSTER CREEK LATENT FATAllTY RISKS WITH OTHER RISKS. +g C . Risk, per yr. ~ LSource-SocTctal .i= in~ifiVi~6a1 N n _ ~, W. L0yster Creek, latent, -0.36 - 2.lE-6=

fatalitiesi

}, 3 Cancer,5U.Sb 321' 1.8E-3'-

Cancer,tN.J.

350! 2E-3 ' ~ ' CancerCU.S!,'malet

45 2.6E-4 m

'l -age ? 25-44. ' ,x Ls M Al'l; Causes~, U.S. 1534 '8.78E-3 a-9, 1 ~ 1 . g x:The f'irstl column =in Table' A-ll gives the source of risk and includes early 4 d@+ c, g - # 4 fatal:ities from~0yster Creek, accidental death (U.S. average.for 1980), and: $fp.f. l death! f rom all causes' (U.S. averag'e. for 1980). The sec'ond column provides-i

the Js'ocietalRrlist from these. 'causes applied -to the ' exposed population around.

y M, i toyster Creek, sin = other words, of th'e '.174,673 persons residing within 20: miles a

j

< off0yster< Creek, Jab'out :82 persons would be expected to die annually from acci- '? Edentst and11,5341 from all causes; The last column provides the= individual risk' ~ .Miichfis obtaihed by dividino the second columr. results by the exposed l population J s('l 74,673 ). ;The table ; indicates t~nat risks from severe accidents imposed by the j J NA' ..s g g {: m' i a + m.,,,, <1s. p -

4y yw* + g3- . g' s' Mi; 1Q f, W %Wy?hF, j4 - p j - 4: >1 m. 4:4- - ' ~ m gg

operation of;thec0yster; Creek. reactor are exceedingly low compared to the

/, [ - 'other risks. iThel.0ysterLCreek' risks are some 5" orders of magnitudesless? I,. (than risksJ from o.ther ' accident's',Jand more than 6 orders' of magnitudelless ~ $ than~ risk'si-[f rom JallY causes. - -m 4 e s - Similarfinformation isj presentedii.niTable' A-12.for latent fatalitles fromi. e - W -.c

0yster; Creek'and/other risks'.. Cancer risks wereiselected as' the primaryi lbisisjfor comparison, since the iatent; fatalities calculated from'Oysterc f Creeklaccidents Jare all ca'ncer ' deaths induced by ' radiation damage <to-cell _s'.-

In!theselcomparisonsEtheiaverage U.S.. cancer rate-(for-1980) was used, as- -;welllas the New Jersey) rate which is somewhat higher. Since susceptibility.

tolfatal.. cancer -is age and-serdependent, the age -' group and' sex with the leastLrisk were also used (the Ref. 5 report did not provide data for age

droups less Lthan 25 Lyears old).. LAs shown in Table A-12,. th'e risk's from Oyster Creek are quite low, being ~ ?sl'ightly)less than three orders of magnitude less than the8U.S. rate,-and: sl'ightly greater than three-) orders Lof magnitudeless.;than(the NewLJersey rate? Tor males, age 25 toi 44, the Oyster Creek ; estimated risk is somewhat - L lessithangl% of the cancer risk for this group. f I ( p M gr9/; w 1 6' "! ) '9 -b i I k ' i . L. -.. 2." -lQ

w. -

-= wa -c yr ~ ;; + y },. h.[ y NQ. ) i n

IREFERENCES:

' ~ - 2 4~ a '--[' 6,il! Jlntedm ReliabilitycEkaluation. Program: Analysis of the Millstone' Point' S ~UnitTNuclear Power Plant, NUREG/CR-3085, May11~983. g~ m ^' a 2.- Reactor Safety Study!- An Assessment of Accident; Risks 'in U.S. Commercial __ a g-

}4uclear Power Plants, @Slf-T4dO70$RRC, Octoberc11i75.

i H3.>

Radio'nuclide' Release Under Spe_cific LWR Accident Condit' ions -' BWR Marktl;

- Design (V5T.~1T, ~BMl!2Td47 J. WGieseke, et.al.,, BatteTTe~ Columbus - Titi6ratories -- 84'(Draft).. I' i4;: LTechnical;GuidanceLfor: Siting Criteria Development,.NUREG/CR-2239,;D.1C. AfdHch, et.af.75andia-National:LabsIllovember lli82. F 5.-- . An Assessment-~of th'e Risk Significance of SEP issuesifor OysterLCreek, P. ; R. Davis, Final: Draf t. Report,. August 1984. 1 i6. ' Calculations of-Reackor-Accident Coeisequences, Version 2 '(CRAC2):. Computer 4 Co'Js~0isFT~GiTTde, tiURTG70i-237CT."T.lftcliie, et.al., $sndlistl abs.,: to i 7 r; g-g l s ^7. _ CRAC2, Calculation of Reactor Accident Consequences, Version 2, Model~ Veic7iTfion, tidRM/CRT25!i2, C~T. _Ritclii~e, et.al., SandTa3T6s.,c to be ; -~ liiiEfii1Td.- ' p. ~8.; LTechnical-Basis. for' Estimating Fission Product -Behavior During~ LWR' Accidents, p' liURIG 6Tif,~liSMC,= June TMl-T x: J - 9.' Statistical-Abstracts'ofothe United States - 1984,104th. Edition,L U.S. Dept. Jh Tof Commerce, -Bureau of the Census. m . u f, ' h10. - Personal' communication,. Paul Volligue, Science Applications ' International, cDecember 114, 1984. 4 > w_., '_. t 1 k.( h .c _I.] 4 M 'f. --f h h f.l.> . (:-' "'J.l4 f, .. v Y h ..'~ \\ s 3 /$ V,. r r..- O m,,i : a,x ,3 I

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