ML20207B676

Draft Technical Evaluation of Emergency Planning Zone Sensitivity Study for Seabrook

ML20207B676
Person / Time
Site:	Seabrook
Issue date:	12/05/1986
From:	BROOKHAVEN NATIONAL LABORATORY
To:	NRC
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ML20207B673	List: ML20207B667 ML20207B676 ML20207J910
References
CON-FIN-A-3852 NUDOCS 8612170316
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L r TECHNICAL REPORT A-3952 1

TECHNICAL EVALUATIPM 0F THE EPZ SENSITIVITY STUDY FOR SEABROOK DRAFT December 5, 1986 Prepared by Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 4

Prepared for U.S. Nuclear Regulatory Commission Washington, DC 20555 Under Contract No. DE-AC02-76CH00016 FIN A-3852

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s o LIST OF CONTRIBUTORS ,

Contributor Affiliation K. Bandyopadhyay Structures & Components Evaluation Group / SAD ,

P. Bezier Civil & Structural Mechanics Group / SAD G. Bozoki Risk Evaluation Group /SRED T-L. Chu Risk Evaluation Group /SRED M. Chun Accident Analysis Group /SRED C. Hofnayer Structures & Components Evaluation Group / SAD M. Khatib-Rahbar Accident Analysis Group /SRED B. Luckas Engineering Analysis & Human Factors Group /ETD J. Pires Structures & Components Evaluation Group / SAD W. T. Pratt SRED A. Tingle Accident Analysis Group /SRED R. Youngblood Facilities Risk Analysis Group P. C. Wang Civil & Structural Mechanics Group / SAD SPED = Safety and Risk Evaluation Division ETD = Enrineering Technology Division SAD = Structural Analysis Division l

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ABSTRACT A technical evaluation of the Seabrook Station Emergency Planning Sensi-tivity Study (PLG-0465) has been performed. This was an evaluation which focused on those areas found to be the most influential in calculating the Seabrook risk estimates. The approach taken by Brookhaven National Laboratory (BNL) was to perfom sensitivity studies to assess the impact on the results in PLG-0465 of the BNL evaluation of these areas, i

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l TABLE OF CONTENTS Page i

Li st o f Co n t ri b ut o r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i i i l Abstract................................................................ V List of Figuaes......................................................... ix ListofTaD1es..........................................................x1 Preface................................................................. Xiii A c k n o w l e d g em e n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . X V sum 9ary..............e.................................................. xvii

1. INTR 000CTION....................................................... 1-1 1.1 Background.................................................... 1-1 1.2 Scope and Focus of the Review................................. 1-2 1.3 3rga ni za:1 on of the Re port . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.4 References.................................................... 1-4

. ~ 2. S Y STEM E V ALU AT ! 0 N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 I n t e r f a c i n g Sy s t en L 0C A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 2 2.1.1 Ge n e r a 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 2 2.1.2 Other ISL Paths........................................ 2-4 2.1.3 initiator Frequencies.................................. 2-4 2.1.3.1 Check valve failure frequencies............... 2-5 2.1.3.2 Cold leg safety injection path frequency...... 2-8 2.1.3.3 RMR suct i on si de f requency. . . . . . . . . . . . . . . . . . . . 2-10 2.1.4 Operator Actions....................................... 2-11 2.1.5 Break Location......................................... 2-15 2.1.6 Ev ent Tree 0u anti fi cation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-16 2.2 Accicents During Shutdown and Refueling Conditions............ 2-17 2.2.1 Los s o f De c ay He a t Re mova 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-19 2.2.2 Low Terpa rature 0v e rg res suri zation. . . . . . . . . . . . . . . . . . . . . 2-21 2.2.3 Lo s s of Co ol a nt Ac ci ce nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-22 2.3 Induced Steam Generator Tube Rupture (SGTR) . . .. . . ... . . . . . . . . . . 2-23 2.4 Co nt a i nme nt Is ol a ti on Fa 11 u re. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-24 2.5 Sunrary....................................................... 2-25 2.6 References.................................................... 2-28

3. EVALUATION O' CONTAINMENT BEHAV10R................................. 3-1 3.1 C a p a c i ty a t Ge n e r a l Yi el d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 ,

3.2 Beh a vi or at La rge De f o rma t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-6 l 3.3 Ca p a bil i ty of Pe net rati on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-11 1 3.4 Sumra y of St ru ctu ral Fi nd i n g s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-17 3.5 References.................................................... 3-18 l

4 C O NT AI NM E NT EV E NT TR E E R E V I E W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 i 4.1 Sen s i t i vi ty to Cont ai nment Lead s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.2 Sen s i ti vi ty to Cont ai nment Perfo rmance . . . . . . . . . . . . . . . . . . . . . . . . 4-5 4.3 Summary....................................................... 4-7 4.a References.................................................... 4-7 v11 l

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5. R E Y !! W OF SOUR C E TE R M5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 Fi d e l i ty t o W AS H-14 00 Met hoc o1 o gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Credit for Scrubbing of Submerged Releases.................... 51 5.3 S u mr a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 2 5.4 References....................................................5-3

6. S IT E C ON SE QUE NCE M00 E L . . . . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 NUR E G -0 3 9 6 B a s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 1 6.2 Co n s e q u e n c e Mo c e l i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2 6.2.1 Whol e Body Dos e Vs Di sta nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.2.2 Thyroi d Do s e Vs Di s t a nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.2.3 Ri s k of E a rly Fa t a11 t i es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.3 Comparisons of Results........................................ 6-5 6.3.1 R e s ul t s of Se a b rook St udy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 6.4 Se n s i t i vi ty 5t ud i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 6 6.4.1 Sensitivity of Results to Multipuff Release............ 6-7 6.4.2 Sensitivity of Results to BNL Review.... .. ........... .. 6-7 6.5 References.................................................... 6-9 viii

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i LIST OF FIGURES l l

Figure Page 5.1 Compariser of Seabrook Station sensitivity results using WASH-1400 source terr. methodology with background. safety goal individual and RMEPS risk levels................................. xx -

S.2 Comparison of median risk of early fatalities at Seabrcok Station for different emergency planning options................. xxi S.1 r.nnprisnn nf .tnhrnnk Statinn ensults in this study and RMEPS with NUREG-0396 - 200-rem and 50-rem whole body dose plots for no immediate protective actions.................................. xxii S.A Comparison of BNL sensitivity studies with PLG 0465 and NUREG-0396.......................................................xxiii S.5 Comparison of 200 rem and 50 rem dose versus distance curves with contributions from shutdown events.......................... xxiv 5.6 Comparison of 200 rem and 50 rem dose versus distance curves for censervative assunption of no credit for operator recovery of o p e r eq u i pme n t h a t c h . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xx y S.7 Ccmparison of all BNL sensitivity studies with PLG-0465 and NL' REG-0396 (200-ree plots with no immediate protective actions).. xxvi 2.1 Frequency of accumulator check valve leakage events............. 2-30 2.2 Ccmparison of 200 rem and 50 rem dose versus distance curves wi th cont ri buti ons f rom shutdown events . . . . . . . . . . . . . . . . . . . . . . . . . 2-31 2.3 Ceeparison of 200 ren and 50 rem dose versus distance curves fcr conservative assumption of no credit for operator recovery i o f ope n equi pme nt hat c h . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-32 2.4 Ccmparison of BNL sensitivity study related to SGTR with PLG-Da65 and NUREG-0396 (200 rem plots with no imediate p r ot ect i ve a cti on s ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 3 3.1 Centainment building cross-section.............................. 3-19 ,

3.2 Cylinder reinforcement.......................................... 3-20 i 3.3 Cont ai neent fi nit e el ement mod el (NF AP) . . . . . . . . . . . . . . . . . . . . . . . . . 3-21 3.4 Pressure-radial displacement relation for containment........... 3-22 4.1 Corposite containment failure probability distributions for benign f ailure, gross f ailure, and total failure................ 4-8 4.2 Ccmparisen of SNL sensitivity study related to early containment failure with PLG-0465 and NUREG-0396 (200 rem plots with no i medi at e p rot e ct i v e act i on s ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-9 6.1 Components of NUREG-0396 curve as computed by BNL using CRAC2... 610 6.2 Risk of death or exceeding dose levels for $1W as calculated by BNL.......................................................... 6-11 6.3 Risk of death or exceeding dose levels for $2W as calculated by B N L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 12 )

6.a Risk of deaths exceeding dose levels for $6W as calculated by l BNL............................................................. 6-13 1 6.5 Dose versus distance curve for release category $1W from l Seabrook for no imediate protective action with BNL results u s i ng MACC S supe ri mpos ed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-14 1x

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LIST OF FIGURES (Continued)

Page Figure .

6.6 Dose versus distance curve for release category 52W from Setbrook for no immediate protective action with BNL results u si n g MACC S su pe rimpo s ed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-15 6.7 Dose versus distance curve for release category $6W from Seabrook for no inmediate protective action with BNL results 6- 16 u s i n g MA C C S s u p e r i mp o s e d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

6.6 Comp a ri son of MACC S to CRAC2 codes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 17 6.9 Comparison of Bhl sensitivity studies with PLG-0465 and NUREG-0395............................................................

6-18 6.10 Comparison of 200 rem and 50 rem dose versus distance curves with Cont ri bution s f rom shutdown events. . . . . . . . . . . . . . . . . . . . . . . . . 6-19 6.11 Comparison of 200 rem and 50 rem dose versus distance curves fo* conservative assumptior of no credit for operator recevery of open equipment hatch......................................... 6-20 6.12 Comparison of all BNL sensitivity studies with PLG-0465 and NUREG-0395 (200-rem plots with no immediate protective actions). 6 21 1

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LIST OF TABLES Table Page 5.1 Impact of BNL Sensitivity Studies on PLG-0465 Risk Estimates..... xxvii 2.1 Summary of Operating Events. Emergency Core Cooling Systen.

Isolation Check Valves. Leakage Failure Mode.................... 2-34 2.2 Summary of Operating Events. Emergency Core Cooling $ystem. >

! solation Check Valves. " Failure to Close" Failure Mode......... 2-37 2.3 Accumul ator Check Val ve Exposu re Data . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-38 2.4 Statistical Data on Leakage Events of Check Valves to A c c u mu l a t o r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,. . . . . . . . . . 2 3 9 2.5 ISL Results Initially Assigned Plant Damage States.............. 2-40 2.6 Pl a n t Op e r a t i n g Mod e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 41 2.7 Categories of 130 Reported Total DHR System Failures When Required to Operate (Loss of Function) at U.S. PWRs 1976 1983... 2-42 2.8 !cpact of BNL Sensitivity Studies on PLG-0432 Risk Estimates.... 2-43 3.1 Statistics of Rebar Yield Strength for Various 5f res............ 3-23 3.2 Reinforcecent Details of the Containment Cylinder............... 3-24 3.3 Reinf orcement Details of the Containment Dome. ....... . . .. .. ..... 3-25 3.4 Statistics of Concrete Compressive Stren 3-26 3.5 Concrete Properties.....................gth..................... ........................ 3-27 3.6 Ch a *a cteri zati on of Cont a i nment Penetrations. . . . . . . . . . . . . . . . . . . . 3-28 4.1 Comparison of Core Melt Frequencies and Distribution of Release Ty p e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 1 0 4.2 Impact of BNL Sensitivity Studies on PLG-0465 Risk Estimates.... 4-11 5.1 Release Categories for Seabrook Station Based on WASH-1400 S o u r c e T e rm Me t h od ol o gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 4 5.2 Revised C-Mat rix for New source Term Categories. . . . . . . . . . . . . . . . . 5-5 6.1 Sunnary of Release Categories Representin Accicents (f rom the R55) . .. . . . . . . . . . . . . ....................... . .g Hypothetical 6-22 E.2 lepact of BNL Sensitivity Studies on PLG-Da65 Risk Estimates.... 6-23 I

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PREFACE This report describes a technical evaluation of the Seabrook Station Energency Planning Sensitivity Study (PLG-0465). The main objective of this technical evaluation is to assist the NRC in its evaluation of the validity of the conclusions presented in PLG-0465. This is therefore a focused review by Brookhaven hational Laboratory (BNL) of those areas identified in PLG-0465 as being the most influential in calculating the Seabrook risk estimates.

However, regardless of the . conclusions of this focused review, BNL cannot attest to the validity of the overall risk profiles presented in PLG-0465.

This follows fron the observation that the risk estimates in PLG-0465 rely heavily on Seabrook Station Risk Management and Emergency Planning Study (RMEPS) (PLG-0432), which in turn relies on the Seabrook Station Probabilistic Safety Assessnent (SSPSA). Unfortunately, the risk profiles in the SSPSA ,have not been independently reassessed, requantified, and validated, by the NRC staff or their contractors. Similarly, within the scope of the review, BNL has also not validated the accident sequence probability estimates in the

• SSPSA. Tnerefore, because these estimates form the foundation for the updated risk estinates in the RMEPS and ultimately in PLG-0465, BNL has not, and cannot, verify the total risk estimates in PLG-0465. This includes the predicted dose versus distance curves. The current review should therefore be regarded as an evaluation of selected issues related to the potential for a large ea rly release of radioactivity at the Seabrook Station and not a reassessment or validation of the total risk profile.

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1 ACKNOWLEDGEMENTS The authors wish to thank Dr. R. Bari _in the Department of Nuclear Energy at Brookhaven National Laboratory for ~ many discussions, comments, 'and suggestions related to this program.

This work was performed for the Division of PWR Licensing - A, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission. The authors wish to acknowledge . G. Bagchi, S. Long, and W. Lyon for their support and guidance throughout the course of this program.

Lastly, the authors acknowledge the efforts of C. Conrad, A. Costini, E. Gilbert, and D. Votruba in preparing this document for publication, i

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SUMMARY

Tnis report describes a technical evaluation of the Seabrook Station Energency Planning Sensitivity Study (PLG-0465). The results of the PLG-0465 Sensitivity Study are reproduced in Figures S.1-S.3. The principal conclusion of PLG-0465 was that an emergency planning zone (EPZ) at the Seabrook Station of I mile radius or less is more justified in terms of its risk management effectiveness than the current 10-mile EPZ was justified by the results of NUREG-0396. This conclusion was made in PLG-0465 without accounting for any new insights about source terms since WASH-1400. The conclusion was based on the following observations:

The individual risk of early fatalities in the population within 1 mile of the site boundary with no imnediate protective actions is less than the NRC Safety goal (refer to Figure S.1). This individual risk is substantially less when a 1-mile evacuation is assumed.

Tne risk of early fatalities with a 1-mile evacuation is conparable to the WASH-1400 results, which assumed a 25-mile evacuation (refer to Figure S.2). The Seabrook Station results for a 2-mile evacuation are substantially less than those for WASH-1400.

The risk of radiological exposures for 1, 5, 50, and 200-rem whole body doses with no innediate protective actions is less at I mile than the corresponding NUREG-0396 results at 10 miles (refer to Figure 5.3).

The Seacrook study (PLG-0465) identified the following three areas as being the nost influential in calculating the Seabrook risk estinates:

The effectiveness of the Seabrook Station primary containment to either remain intact or to maintain its fission product retention capability for periods much longer than required for even delayed, ad hoc protective actions.

A more realistic assessment of the strength and failure modes of the Seatrook containment than was possible within the state-of-the-art of PRA when the RSS was completed.

A nore realistic treatment of the initiation and progression of inter-facing systens LOCA sequences.

At the recuest of the NRC, the BNL technical evaluation focused on the areas that were identified in PLG-0465. The approach taken by BNL was to per-form sensitivity studies to assess the impact on the results in PLG-0465 of the BNL review of these areas. The BNL sensitivity studies used the condi-tional risk indices provided in PLG-0465 (and supporting documentation) to assess how changes in the probability of accident sequences and containment failure nodes would change the risk estimates in PLG-0465. The sensitivity studies calculated revised 200 rem-dose versus distance curves and the mean absolute risk of early fatalities and total cancer fatalities. The results of the BNL sensitivity studies are given in Table 5.1 and Figure S.4 The dose vs distance curves in Figures S.3 and S.4 can of course be directly compared:

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however, the BNL mean absolute risk numbers in Table S.1 cannot be directly conpared (without additional calculation) with the information from PLG-0465 in Figures S.1 or S.2. Tne nean absolute risk nunber in Table S.1 would have to be converted into individual risk of fatalities in the population within 1 nile of the site boundary before direct comparison within information in Figure S.1 could be made. However, the information provided in Table S.1 is a useful indication of how the PLG-0465 risk levels in Figures S.1 and 5.2 would change if recalculated using the assumptions of the sensitivity studies.

In addition to the areas identified in PLG-0455, other areas were identified as potentially inportant to risk at Seabrook. In particular, the applicant was requested to provide information on the risk associated with accidents during shutdown. The results of the applicant's assessment of such accidents are also given in Table S.1 and in Figures S.5 and S.6. BNL was not able to assess the frequency of these events because there remain fundamental questions regarding modeling of these scenarios. In addition, the potential for induced stean generator tube rupture (SGTR) for accidents in which the prinary systen is at high pressure was identified as a topic for review. This topic was reviewed in detail by the NRC staff and is the subject of continuing NRC and industry research activities. In an effort to assess the potential influence of SGTR, a simple sensitivity study was performed at BNL and the results are also given in Table S.1 and Figure S.4..

In Table S.1 and Figure S.7 the effect on risk of combining all the sensitivity is presented. This calculation should not be interpreted as a reassessnent of the ovarall risk profiles for Seabrook because there was not a systenatic attenp by BNL to obtain completeness. It is simply intended to indicate how :ne results of the various sensitivity studies could influence the risk estinates in PLG-0465. The method used to combine the effect of all of tne studies is not rigorous and could lead to inconsistencies. In addition, it is not normal practice in probabilistic risk assessments to conbine bounding sensitivity studies. The results in Table 5.1 and Figure S.7 should therefore be recognized for what they are, namely, a series of sensitivity studies and not be interpreted as a statement of the overall risk at Seabrook.

The results in Table S.1 and Figures S.4-S.7 are useful to focus on those areas of the BNL review that appear to have the greatest impact on the conclusions in PLG-0465. Tne results indicate, given the extent of this focused review, that the conservative assumptions regarding accidents during shutdown and induced SGTR have the nost impact on the dose vs distance and j risk estinates in PLG-0465. However, the more optimistic assumptions regard- l ing these events have ninor impact on the PLG-0465 results. These are areas 1 of considerable uncertainty. which at the present time do not allow a better {

definition of the risk estimates than given by the ranges in Table 5.1 and Figures S.4-5.7 The conclusion of this focused review isi that the risk estimates quoted fron PLG-0465 at the beginning of this summary do appear to be influenced by the various sensitivity studies performed at BNL. The individual risk of early fatalities with no immediate protective actions as predicted in PLG-0465 are only slightly below the proposed NRC safety goals as shown in Figure 5.1.

Theref ore, fron an inspection of Table 5.1 it is clear that some of the sensi-tivity studies will result in risk estimates (if no immediate protective xviii

• e M R T*-[*kf actions are taken) that approach and rey exceed the pr;p.;ad goali" However, with a 1 mile evacuation none of the selected sensitivity studies exceeded the safety goal, in addition, the statement in PLG-0465 that "there is no significant frequency of exceeding 200 rem beyond 1.5 miles in the Seabrook sensitivity results" is not confirmed by some of the sensitivity study results presented in Figures S.4-5.7.

Whether or not the results of the BNL (and applicant) sensitivity studies and the f act that BNL cannot verify the total risk estimates at Seabrook chal-lenge the principal conclusion of PLG-0465, namely "that an emergency planning zone (EPZ) at the Seabrook Station of 1 mile radius or less is more justified in terns of its risk management ef fectiveness than the current 10-mile EPZ was justified by the results of NUREG-0396" is a matter of NRC policy and beyond the scope of the BNL review.

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10'I b BACKGROUND ACCIDENTAL FATALITY RISK E 10-3 _

(5 FATALITIES PER 10,000 POPULATION PER YEAR)

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WASH-1400 source term methodology with background, safety '

coal individual and RMEPS risk levels.

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0 10 1C' 10 2 10 3 4 5 10 10 EARLY FATALITIES Figure 5.2 Comparison of median risk of early fatalities at Seatrook station for different emergency planning o pt i or.s . (Reproduced from PLG-0465, April 1986) xxi

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Figure S.3 Comparison of Seabrook Station results in this study and RMEPS with fiUREG-0396 - 200-rem and 50-rem whole body dose plots for no immediate protective actions. (Reproduced from PLG-0465, April 1986) xxii

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(Calculations performed by applicant) s' xxiv

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XXV

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l l

xxvi j i i i

l

Table 5.1 Inpact of BNL Sensitivity Studies on PLG-0465 Risk Estimates Absolute Risk Per Reactor Year Early Fatalities Total Cancers Sensitivity One Two One Two No Mile Mile No. Mile Mile Study Evac. Evac. Evac. Evac. Evac. Evac.

Original 2.7(-3)* 3.6(-4) 2.4(-5) 1.5(-2) 1.4(-2) 9.2(-3)

PLG-0465 Results Revised 2.8(-3) 4.6(-4) 1.2(-4) 1.5(-2) 1.4(-2) 9.3(-3)

Frequency for Inter-facing System LOCAs i

Stean **

4.5(-3) 2.2(-3) 1.8(-3) 1.7(-2) 1.6(-2) 1.1(-2)

Generator Tube 2.8(-3) 4.2(-4) 8.4(-4) 1.5(-2) 1.4(-2) 9.3(-3)

Ruptured Containment 4.7(-3) 2.4(-3) 2.0(-3) 1.8(-2) 1.7(-2) 1.2(-2)

Loads anc j Performance Accidents 6.2(-3) 2.9(-3) 3.0(-4) 2.1(-2) 2.0(-2) 1.4(-2) fron Shut- ***

2.7(-3) 3.8(-4) 2.8(-5) 1.5(-2) 1.4(-2) 9.2(-3) down****

Irpact 1.0(-2) 7.0(-3) 4.2(-3) 2.6(-2) 2.5(-2) 1.9(-2) of all .

Issues ***

4.9(-2 2.6(-3) 2.2(-3) 1.8(-2) 1.7(-2) 1.2(-2)

• 2.7(-3) = 2.7x10-3

• • Pessimistic assumptions.

• • • 0ptimistic assumptions -

• • • • Calculated by the Applicant, not confirmed by BNL.

] sxvii

1-1

1. INTRODUCTION 1.1 Backgrounc The Seabrook Station Probabilistic Safety Assess' ment (SSPSA)1 was com-pleted by Pickard, Lowe, and Garrick (PLG) Inc., for the Public Service Company of hew Hampshire and Yankee Atomic Electric Company in December 1983 and subnitted to the Nuclear Regulatory Commission (NRC). The NRC staff and its supporting contractor initiated an in-depth review of sections of the SSPSA related to determining those accident sequences that could lead to core damage. However, this effort was terminated prior to completion of the review.2 A se;:arate contract was placed with Brookhaven National Laboratory (BNL) to perform a very limited 3review of those portions of the SSPSA related to core reltdown phenomenology, containment response, and radiological source terms. The BNL review 3 did not include an assessment of the physical strength of the Sea ock Containment. .

The key results of the SSPSAI are given below:

. Tne reaa and redian values of the uncertainty distribution for core nelt frequen:y were found to be 2.3x10 " and 1.9x10 " events per reacto -year, respectively.

. Botn the societal and individual risk provisions of the NRC safety goals were ret by wide nargins; hence, the risk to public health and safety was estinated to be extremely small.

. Different risk factors were found to have different key contributors.

Interfacing systems LOCA events and, to a lesser extent, seismic-induced transient events were the principal contributors to early health risk. The contributors to core melt frequency and latent health risk were made up of a large group of initiators, including loss of offsite power, transient events, fires, and seismic events.

. The dominant contributors to core melt frequency were support system f aults, external events, and internal hazards that af fected both the

1-2 core cooling and containnent heat removal systems. As a result, a najor fraction of the core melt f requency, 73%, was associated with sequences in which long-term containment overpressurization was indi-cated, while only It was associated with early containment failure.

In contrast with previous containment analysis, the timing of contain-ment overpressurization in the above sequences was found to be neasured in units of days rather than hours.

A major result of the SSPSA was that interfacing system LOCA events were the principal contributors to early health risk. The results of the SSPSA were updated in the Seabrook Station Risk Managenent and Emergency Planning Study (RMEPS), PLG-0432" to account for - new insights regarding radioactive release source terms and the progression of sequences involving loss of coolant events that bypass the containnent.

i The purpose of tne RMEPS was to present the results of a technical evaluation of emergency planning options and other risk management actions

$ that were under consideration for the Seabrook Station. The principal focus of the study was the evaluation of the impact of various protective actions such as evacuation and sneltering to various radial distances from the plant t

site.

A second report related to emergency planning was published (Seabrook Station Energency Planning Sensitivity Study, PLG-0465) which determined the radius of the EPZ that could be justified without consideration of any advances regarding the source term methodology since the completion of the Reactor Safety Study in 1975. It is this second study that is the focus of the current BNL review.

1.2 Scope and Focus of the Review The objective of this technical evaluation report is to assist the NRC in 1 l

evaluating the robustness of the applicant's conclusions regarding the Erercency Planning Sensitivity Study for Seabrook (PLG-0465). The focus of i

1

1-3 the BNL review reflects those areas of PLG-0465 (and the supporting docunent PLG-0 32) where najor risk reductions (when compared with the results of the SSPSA) were calculated. Thus, our review assessed the physical strength of the Seabrook containment and the magnitude of the challenges to it. In addition, the potential for bypassing the containment via interfacing systen LOCAs or by loss of containment isolation was assessed.

An inportant comparison in PLG-0465 relates to a comparison of the dose and distance risk curves developed for Seabrook with similar curves in NUREG-0396. Thus, the technical bases for the risk curves in NUREG-0396 will be reviewed and compared with the bases for the curves presented in PLG-0465 to deternine the appropriateress of the conparison. Finally, the site conse-quence modeling in PLG-0465 will be evaluated and compared with current NRC consequence nodels.

1.3 Organization of the Report The previous section identified the focus of our review of PLG-0465 and indicated the linitations of the effort. The report is organized to address each of the areas discussed in Section 1.2. Initially, in Section 2, those portions of PLG-0465 and the RMEPS (PLG-0432) related to system failure are reviewed to determine the appropriateness of the accident sequence probabili-ties.

Section 3 reviews the ability of the Seabrook Station primary containnent to withstand the very severe pressure / temperature loads associated with core neltdown accidents. This is a very important review because the applicant considers that the Seabrook containment has a significantly greater capability for containing core meltdown accidents than a number of other large dry con-tainrents that have been reviewed by the NRC staff over the last several years.

4 The sensitivity of the conclusions in PLG-0432 to uncertainties in containnent loads (pressure /tenperatures histories) and containment perfor-nance (based on the review in Section 3) is explorr.d in Section 4 The source terns used in PLG-0432, which were based on RSS methodology, are reviewed in  ;

1-4 Section 5. Finally, in Section 6, the site consequence model and the risk calculations presented in PLG-0432 are reviewed.

1.4 References

1. SSPSA.

2. LLNL Review Craft.

3. NUREG/CR-4540.

I

o e 2-1

2. SYSTEM EVALUATION In this section, those positions of PLG-0465 and 1 the RMEPS2 (PLG--043?)

related to systen failure are reviewed to determine the appro-priateness of the accident sequence probabilities. This review focused on the reduced frequency of the interfacing systen LOCA claimed in PLG-0432. This section also addresses other sequences that might lead to early containment bypass and hence the potential for a large early release of radioactivity. This is a particularly inpartant topic for review because the applicant's risk estimates are based upon the conclusion that essentially all accidents that may lead to a large early release of radioactivity have a very low probability (less than one chance in ten nillion per year of reactor operation).

At NRC direction, since the present concern is with the frequency of a substantial release, this section addresses several selected areas which are alreacy known to bear directly on this question. The strength of. containment is addressed in a later section; here, selected modes of containment bypass are discussed. These include classical interfacing systens LOCA, f ailure of containment isolation, and breakdown of steam generator tube integrity during a seve'e at:icert.

The study generally creates a strong impression of completeness.

However, the presert decision logic imposes unusual . demands on the analysis.

Wnen ore is quantifying a multiple passive failure event whose frequency and consequences are both clearly dominated by events whose absolute likelihood has been established experientially, some latitude can be tolerated in the cuantification, and neglect of the less likely contributors can be justified.

Here, the decision process requires that certain categories of events be shown to be essentially incredible on an absolute scale; there can be no dominating events for then to hide behind.

Several types of questions arise:

1) whether the basic assumptions of the submittal are adequate, and whether the analysis confronts all the important issues r-- -

a 0 2-2

2) whether, given the basic assumptions, the plant-specific modeling is self-consistent and complete within the range of issues addressed by the nodeling

3) whether the scenarios actually modelled are properly quantified (including common cause considerations).-

An example of a Type 1 completeness question will be discussed here in Section 2.3, narely, whether a high-pressure melt scenario can lead to steam generator tube degradation and concomitant containment bypass. This is a fairly generic question, and one that is not answered within the material sub-mitted to date. This question is presently being studied by the utility and i*s consultan*s.

An exarple of a Type 2 completeness question is whether the check valve between residual heat renoval (RHR) suction and the refueling water storage tank (RWST) is likely to f ail in scenarios wherein the RHR system is overpres-surized. Tne study raises and discusses this question, but without clearly establishing why the check valve survival probability is high enough to warrant not exploring such scenarios. We believe that this information will be feathcoming, but while this particular question may be laid to rest, it is entleratic o' a f amily of such auestions which cannot be exhaustively tallied within the scope of this limited review, but which bears on the subnittal's conclusiers.

Finally, a particular instance of a type 3 question will I be discussed be l o,: in Section 2.1, where the frequency of multiple check valve failure will l l

be addressed.

2.1 Interfacire System LOCA 2.1.1 General According to the Seabrook Station Emergency Planning Sensitivity Study,I one of the principal contributors dominating early health risk--and one which has been subjected to extensive reanalysis since the SSPSA--is an Interf acing Syste s LOCA that bypasses containnent. From all the potential pathways

2-3 through which an Interf acing Systems LOCA (ISL) may occur, the study identi-fied six lines as possible initiators for ISLs:

. Four lines in the cold leg safety injection (Low Pressure Injection

[LPI]/ Residual Heat Removal (RHR] Loop Return lines)

. Two lines in the suction side of the RHR system.

The corresponding initiator frequencies were determined as:

Cold Leg Safety injection Path (VI): 4.5x10 6 event / reactor year RH: Suctio- Side Path (VS): 3.3x10 6 event / reactor year which resulted in the following core damage frequency: ,

CDISL = 4.4x10-8 event / reactor year.

These'f recuencies were cbtained as a result of an enhanced and innova-tive analysis which involved new treatments of various aspects of the acci-dent. The new treatments are listed below:

More conplete modeling of valve failure modes New data on check valve failures versus leak size More realistic treatnent of dynamic pressure pulse

. Explicit rodeling of RHR relief valves Quantification of RHR piping fragilities to overpressure j Modeling of RHR pump seal leakage I

. Operator actions to prevent melt c;fW.ccre Tnernal hydraulic and source tern Letor .JJeled using MAAP"

. Uncertainties quantified.

1 BNL has perfe med a limited review of the new analysis, compiled a number l of questions and observations, and performed a new evaluation of the initiator frequencies. Tne objective of this chapter is to provide the results of this limited review.

o- a 74 2.1.2 Otner ISL Paths BNL perfomed a cursory line survey for potential containment bypassing pathways for ISL. The survey identified some pathways (e.g., RHR lines to the RCS hot legs, letdown line, excess letdown line, etc.) which were ignored in the Seabrook study. Since BNL believes there were good reasons to ignore then, these pa':hways were not analyzed. However, BNL believes that all the potential paths should have been considered and the reasons for the rejection of each path, with the probability of the path being open, should have been discussed.

2.1.3 Initiator Frecuencies The detemination of the initiator frequencies is one of the most impor-

~

tant parts of the Seabrook EPZ Study.1 It depends' essentially on the correct estinate of the frequencies of relevant failure modes of valves in various interfacing lines. Tnese valve failure modes are:

Disc rupture or gross leakage of series valves (check valves) in the LDI lines Disc rupture or gross leakage of MOVs, failure of sten mounted limit switches, and disc f ailing open when indicating closed, in the RHR suction lines.

Tre approach applied for mooeling of initiator frequencies in thc Scaorook stucy is based on two " innovative" steps:

a) Separation of the gross (reverse) leakage of check valves f ailure mode into " gross reverse leakage" and " failure to reseat on demand" failure modes, which were treated together in earlier data bases. This step i required, of course, a "remodeling" of the appropriate Boolean equations, b) An analysis of data on check valve leakage frequency versus leak rate for check valves ~ of the RCS/ECCS systen boundary. This step resulted in applying a reduced check valve leakage failure frequency in the quantifi-cation of the initiator models.

1 .

o .

1 2-5 In the process of surveying data of the Nuclear Power Experience (NPE) data base, no disc rupture events were identified by PLG for check valves and N3Vs. Tne maximum observed leak rate was 200 gpn. Leak rates were estinated based on other available evidence: the rate of boron concentration change in the accunulators, pressure reduction, and similarity to other occurrences for which the leak rates were known. To estimate the total check valve hours, the information provided in NUREG/CR-13635 on the number of valves in the ECCS in various PWRs was used. PLG's total number of check valve-hours was approxi mately 1.0x108 . To estimate the frequency of check valve failure to reseat on demand, two types of data were used: estimates from generic sources of f ail-ure data, and experiential data from eight U.S. nuclear plants for which PLG perfornec plant specific PRAs.

2.1.3.1 Check valve failure f.requencies Since the check valve f ailure frequencies play a crucial role in the ISL analysis, BNL perforned a somewhat more detailed review of that part of the Seabrook study. As a consequence of the review process the following observa-tions are made:

a' PLG selected a particular subset of those events listed in the NPE data base, nanely, events involving check vsives at the RCS-ECCS interface.

t} To estinate the total number of check valve hours, it used the total popu-lation of check valves in the ECCS instead of the particular subset of check valves at the interfaces. This resulted in substantial overestima-tion of check valve hours, c) The correct exposure time for check valve failures is not merely the time when the plant is operating. Fcr example, check valves in the RHR are ,

almost continuously exposed to potentially degrading conditions (during cold shutdowns, as well). A correction factor for pressure exposure of l

interfacing lines should be considered separately, in calculating the initiator frequencies. ,

d) When estimating leak rates for accumulator check valves fron accunulator l inleakages, it must be recognized that the deduced leak rates relate to two check valves in series, rather than lealage through a single check valve.

2-6 e) The leak failure frequencies versus leak rate curve presented in'the study (reproduced in Figure 2.1) is only a first approximation for a more pre-cise leak f ailure frequency versus relative leak rate curve. In particu-lar, this curve pooled data involving a variety of check valve sizes. A .

more sophisticated treatment would require knowledge of the size popula-tion of check valves at the interfacing pathways, f) The largest leak rate in Fig'ure 2.1 is of the order of 200 gpm, whereas the arena of interest ranges to 65,000 gpm. The " linear" extrapolation to higher rates is not necessarily justified. If the shape of the distribu- I tion is Pareto, the linear extrapolation is in order. However, if it L

follows a Rayleigh distribution, the extrapolation is not correct (but conservative). Seabrook-specific considerations (valve sizes, designs) are not nade in the analysis. ,

g) The initiator nodels implicitly assume that the leak tests of the valves

" discover" all failures and valves behave as 'new after each test. The I stucy does not describe the relevant test processes and the expected "real" efficiency of these tests. I h) The report does not cons' der conmon cause failures. Such failures indeed ,

happen daa to boron deposition, improper neintenance such as installation of inprcper conponents (gaskets, seats, or valve disks) which may fail almost irreciately or at a later time.

4 In order to estimate quantitatively the consequences of some of the above centionec deficiencies, BNL performed a more detailed reevaluation of relevant check valve failure data. The process was facilitated by the availability of

}

relevant f ailure events selected for an independent study of ISL at PWRs, t which is presently ongoing at BNL for the NRC.

i t

Tables 2.1 and 2.2 present failure events for High Pressure / Low Pressure t isolation check valves selected by BNL. These tables contain more relevant events than are listed in Table 3.8 of PLG-0432.

• 1 Table 2.1 contains events for the valve " leakage" failure node. Table 2.2 presents the events for the " valve failure to close" failure mode. Table 2.1 shows also data on the estinated leak flow rates. These latter data are 03 ainet essentially with the sane nethod as those of Table 3.8 of PLG-0*32. I U

b &

2-7 ilsing the failure events related to the accumulator check valves only, BNL attempted to determine the leakage failure rate for 4. " clean" subset of check valves. The appropriate check valve population was relatively easy to determine for all the PWRs in the U.S. (see Table 2.3). The total time fron start of conmercial operation of the individual plants was used as " time of exposure" for this clean subset of check valves, since, e.g..' water with boric acid degrades these valves. The total number of check valve-hours obtained is 2.34x107 .

The frequencies of accumulator check valve leakage events for various leak rate ranges are given in Table 2.4 The corresponding frequency exceed-ance/hr values are plotted against the check valve leak rates in Figure 2.1.

For comparison, Figure 2.1 shows also the PLG data. The shape of the curve is almost identical with that of PLG, but shifted higher, with almost one order of magnitude, due to the higher number of events selected and the more precise value for check-valve-hours.

It is appropriate to nention here several precautions concerning the leakage failure characteristics derived fron accumulator check valve failure events.

In the NPE data base the majority of interfacing check valve leakage events involve accumulator valves. Although this seeming bias could arise from the extra monitoring of the accumulator, it could also reflect a particularly severe environment acting on the valves. If i the latter is true, then leakage exceedance frequency data (ordinates i l

in Figure 2.1) nay lead to overestimates of the frequency for other 1 interfacing check valves.

The leak flow rate data'(" leak sizes"; abscissas in Figure 2.1) repre-sent lower limits for these quantities, because leakage flow rates i estinated from accunulator inleakages involve, in most of the cases, leakage through two check valves in series, where the less-leaking valve dominates (the other valve may be even wide open).-

As a result of these factors, a more realistic leakage failure exceed-ante frequency /hr versus leak rate curve for non accunulator interfac-ing paths may be somewhat lower in frequency at low leak rates, but

o e 2-8 might fall off more slowly with increasing leak rate than do the curves in Figure 2.1.

. This more realistic curve yet has to be determined.

Since there are no nore accurate data available, BNL recalculated the initia-tor frequencies by using the data obtained for accumulator check valves.

Since the purpose of this calculation is to contrast the result with that of the PLG analysis, the sane extrapolation and calculational techniques are used as those of PLG.

2.1.3.2 Cold leg safety injection path frequency This section presents a revised estimate of VI, the frequency of inter-facing LOCA through the injection lines. The calculation presented below is intended to follow the PLG analysis step by step, except that the check valve failure statistics have been nodified as indicated above. Subsequently, these nodified initiator frequencies will be propagated through to illustrate new plant darage state f requencies. However, it should be noted that one problen noted above is not addressed by this modification: the fact that much check valve leakage experience actually corresponds to leakage through nultiple checi valves. For example, referring to Figure 2.1, if we take 10 8 per hour as the frecuency of 1800 gpm leakage through multiple check valves, we obtain an annual frecuency on the order of 10 per year (per line) for these rather sizable events. At 10 7 per hour for 150 gpm leakage, we would obtain 10 3 per year, Aich is near the threshold of observability. We belihve that this treatment is sinplistic and conservative, and accordingly have not propagated these numbers; but on the other hand, a proper treatment would ultimately have to confront the nultiple leakage aspect of the accumulator experience, which neither PLG's analysis, nor the one presented below, accomplishes. The following is, then, a recalculation of the PLG result using PLG methods but modifying the single check valve failure rate as previously. discussed.

Fron Figure 2.1, the nedian frequency of a single check valve failure resulting in leakage, that exceeds the capacity of one charging punp (i.e., ,

i 150 gon) is about 1.1x10 7 per hour. Assuning a lognormal distribution for l

2 a 2-9 this frequency and a range factor of 10 (which nay be too conservative for this increased statistic) yield:

Parameter Frequency (events per reactor year) 95th percentile 9.6x10 3 Mean 2.6x10 3 Median 9.6x10 "

5th percentile 9.6x10 5 i

Sinilarly, the rediar. f requency of exceeding 1800 gpm is 1.4x10 e per hour.

Assyning a lognornal distribution with a range factor of 14 yields:

Paraneter Frequency (events per reactor year) 95th percentile 1.7x10-3 Meaa 4.4x10-4 Median 1.2x10 "

Sth percentile 8.8x10-6 The frecuency of "f ail to operate on demand" for check valve, ad is taken to be the sare value as that used by PLG.

Ac = 2.7x10 " (mean value).

By using Fornula 3.14 of PLG-0432, the estimated mean frequency of failure of two series injection check valves, that produces leakage to the RHR systen in excess of 150 gp: is:

4.90x10 s events / reactor-year.

Since there are four injection paths, the mean value for Cold leg Safety Injection Path, VI = 1.96x10

• events / reactor-year.

4

, e 2-10 Top event, LR in the injection path event tree represents the fraction of the initiating event frequency, VI, in which the leakage not only exceeds 150 gpn, but. also exceeds 1800 gpm. The product of LR and VI thus represents the fre-quency of pressure challenges to the RHR system due to failure of both check valves in the four injection paths. Based on the above values, LR has a mean '

value of .058.

2.1.3.3 RHR suction side frequency For an ISL to occur in the RHR hot leg suction path, failure of two series MOVs nust occur. In the PLG-model for this path, the failure involves:

r a) independent failures of both MOV valves, causing excessive leakage; or j b) independent failure of one of the valves and a demand failure of the second valve, or c) " valve f ail open while indicating closed" f ailure for the first value and excessive leakage failure of the second valve.

Ir. the PLG treatrent, the frequency of MOV valve disc leakage and failure upon denanc (due to a sudden pressure loading) were assumed to be identical to that for the creck valves. For the frequency of failure of an MOV to close on demand but indicate closed, a nean value of ad = 1.1x10 " failure / demand was used in the PLG treatment.

Applying the sare approach as PLG (Formula 3,15 of PLG-0432) with the newly deterrined check valve leakage frequency, BNL recalculated the total (2 line:) suction side ISL frequency, VS. The new mean frequency for the RHR suction side path is:

VS = 1.44x10 " events / reactor-year.

The split fraction, LR, for the fraction of VS in which the leakage past the series MOVs is greater than the capacity of the relief valves, is practically the same as in the case of injection lines: .058.

1 I

1 i

~ , - . , - - - - . , - . _ ,

o s 2-11 For these check valve leak rates, ~ the PLG procedure of using the check valve leakage failure rates as " conservative" estimates for the leakage f ailure rates of the MOVs is probably too conservative, and appropriate MOV 4

leakage failure frequencies should be used.

, in the case of a reanalysis of the initiator frequency, VS, there are several BNL observations to be taken into account in a new suction side ISL model:

a) Inadvertent opening of the two MOVs due to common cause failures such as improper raintenance, malfunction of the interlock system, design error, inproper tests, or testing operations, b) Failure of the sten or other internal connections in valves equipped with limit switches or failure of a limit switch (including improper mainte-nance such as reversing indication).

c) It is dif ficult to see why only two MOVs have limit switches, instead of four, di lt woald be very useful to describe the valve inspections that are pro-g nised each time the plant goes to cold shutdown or is refueled. For ,

exa ple, at a plant recently investigated by NRC, Region 1, everything was tested tnoroughly, but the relays for the MOVs were not inspected.

e) Considerations should be given to operating procedures and the likelihood that the procedures will not be followed, and interlock behavior.

?.l.4 Operator Actions 1 . .. ; .

The ahility of the Seabrook operators to diagnose, respond to, and miti-i gate a Reactor Coolant System (RCS) to Residual Heat Removal (RHR) Interfacing Systems LOCA will be reviewed in this subsection. Appropriate operator actions can mitigate the consequences of the ISL sequence that result in leakage outside containment when the capacity of the RHR punp suction relief valves is exceeded and subsequent failure of.,t'he RHR pump seals results. As discussed by PLG-0432, the success of these mitigative actions is dependent on the ability of the Seabrook operating staff based on their training and emer-gency procedures, to correctly diagnose a LOCA outside containrent. Tne

3 ~

2-12 correct diagnosis nay be hampered by operator confusion between symptons asso-ciated with those LOCAs inside containment which fill and pressurize the Pres-surizer Relief Tank (PRT) by pressurizer relief or safety valve discharge flow and those associated with a RCS-RHR Interfacing Systems LOCA outside contain-nent which also fills and pressurizes the PRT via the RHR suction relief valve discharge flow.

The following is the result of a brief BNL evaluation of operator diagno-sis and actions to mitigate the consequences of RHR pump seal failure Inter-facing Systems LOCA sequences at Seabrook and its assessment in PLG-0432,2 Section 3.1.4.3 entitled " Operator Actions and Emergency Procedures."

Tnis evaluation was preceded by an independent and fairly extensive familiarization preparation with the Seabrook procedurcs as they relate to the ISL to be studied. This preparation was followed by observation of a series of Seabrook Sinulator demonstrated accident sequences which illustrated the distinguishing characteristics of the LOCA outside containnent and the responses expected of the Seabrook operators. The BNL evaluation was per-forned by a forrer Senior Licensed Operator and Westinghouse Reactor Plant .

sinulator Ceatified Engineer. A nore detailed and complete evaluation of

~

operator res;o se would require a conprehensive Hunan Reliability Analysis (HRA) such as Tean Enhanced Evaluation Method (TEEM)6 by a knowledgeable team of specialistis providing expertise in PWR operations, PWR systens engineering and hunan reliability. This team would develop a detailed task sequence anal-ysis of tre Seabrook operating staff performing the detailed tasks required to nitigate these sequences and analyze the associated human reliability of the staff response using the analysis.

There are three sets of operator tasks identified by PLG-04322 which are to be inportant to the mitigation of the sequences by the Seabrook operating staff (each with a unique Operator-Action Sequence identification number in parenthesis), namely:

(01) Diagnose the RHR system LOCA (02) Isolate the RHR systen LOCA (03) provide nakeup to the RWST.

2

• 2-13 To successfully accomplish these tasks, the operating staff must follow the appropriate parts of the following Seabrook procedures which are appli-cable to the RHR systen LOCA event.

. Procedure E-0 (Reactor Trip or Safety Injection), Rev. 00, dated 05/16/86. -

. Procedure ECA-1.2 (LOCA Outside Containment), Rev. 00, dated 05/16/86. This procedure provides guidance on isolating the rupture.

. Procedure ECA-1.1 (Loss of Emergency Coolant Reci rcul ati on--ECR ),

Rev. 00, dated 05/16/86. This procedure provides guidance for supply-ing adeauate ECCS flow and plant stabilization.

. Procedure E-1 (Loss of Reactor or Secondary Coolant), Rev. 00, dated 05/16/86. This procedure provides guidance for long-term cooling and stabilization.

Please note that ECA-1.2, Rev 00 needs to be revised to ensure that valves PH-V21 and -V22 the RHR pun;;, hot leg injection cross connection valves are closed prior to trying to identify and isolate a break in the low pressure systems. Tnis need was identified by a detailed review of the above four prececures.

Tne quantification of the three operator tasks " identified by the Operator Action Sequence identification numbers 01, 02, and 03 above have been provided ir PLG-0 32,1 Table 3-10. According to the accompanying discussion in Section 3.1.4.3, "Tnese operator actions include the hardware contribution, where applicable, and are based on enhanced procedures and instrumentation in order to aid the operators in their diagnosis of the event." For each of the three operator tasks, a " base," human error rate with a "mean" value of 0.005 has been identi- fied as "0P".

This singular human reliability analysis number is identified in PLG-0432 2 as ". .

. recomended in Table 20-6 of NUREG/CR-12787 . . . for following a procedure under abnormal conditions. This human error rate is interpreted to have a nean value of 0.005 and to be represented by a lognornal distribution range f actor of 10." Therefore, the only part of the tnree sets l

l

e o 2-14 of operator tasks 01, 02, and 03 which changes their quantifications values is the hardware contribution. The human reliability quantification contribution of each of three operator tasks use the same estimated Human Error Reliability (HEP) value of 0.005 with an error factor of 10. Each HEP is based on NUREG/CR-1278,7 Table 20-6 (entitled " Estimated Human Error Probability (HEP) related to f ailure of administrative control"). Item (4) HEP entitled "Use written operations procedures under abnormal operating conditions." There-fore, no differentiation is made to distinguish quantitatively among operator actions related to " diagnose " to " isolate," and to " provide."

The October 15, 1986 demonstration at the Seabrook Simulator with the abovenentioned Seabrook abnornal/energency related procedures provided (in the absence of a detailed TEEM' equivalent human reliability analysis performed on an actual Seabrook licensed operator shift) a reasonable assurance that a licensed Seabrook crew would adequately perform the necessary actions within the tire required on the sinulator. This assurance is heightened especially since the Seabrook Traininc Center has recently instituted, in October 1986, a training rodule entitled "RHR Interface LOCA/ Student Handout," as part of its Requalification Training Progran. The inclusion of such a module will rein-force the inportance of the RCS-RHR Interfacing Systens LOCA. Please note that this nodule confirrs the need to revise Seabrook Procedure No. ECA-1.2, Revi-sion No. 00, dated 5/16/86 entitled "LOCA Outside Containment" to close (or verify closed) valves RH-V21 and -V22, the RHR punp hot leg injection cross connection valves to identify and isolate a break in the low pressure systens.

Nevertheless, there were a number of concerns raised during a plant walk-through 09 the sane date as the simulator demonstration which the Seabrook Sinulator cannot adequately answer. These concerns include the following:

a) Ability of RHR pump leakage to be detected in the control room - con-cern lies with vault compartmentation design with the Equipnent Vault sump not receiving leakage promptly therehy delaying level detection input in the control roon, b) Ability of RHR purp relief discharge into the PRT to be distinguish-able in the control roon from the pressurizer relief and safety valve I

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2-15 discharge - concern with the latter relief and safety valve dis-charge tailpipe tenperatures.

In sunnary, the operator action analysis performed in PLG-0432.1 Section 3.1.4.3 appears to be superficial at best. The use of one single HEP value 7

f ron one table of NUREG-1278 is an example of a lack of detailed and insuffi-cient task analysis in assessing human performance appropriately. A detailed 6

TEEM equivalent human reliability analysis is a far more appropriate and rigorous approach to assessing Seabrook operator actions during a RCS-RHR Interf acing systens LOCA. Nevertheless, the simulator demonstration empha-i sizing hunan reliability and task sequence timing, plus the new training nodule reinforcing a conmitment to train the operator, provide a reasonable assurance that a licensed and trained Seabrook crew would adequately perform the necessary actions within the time frame required.

2.1.5 Break Location Tne " weakest link" of the RHR pressure boundary when subjected to acci-dental pressurization was identified by the applicant to be the RHP pump i seals. A tabular listing of f ailure probabilities at 2250 psia showing pump seal failure proDabilities ranging to 0.5 while metallic failure probabilities (piping, valves, and tubing) were 0.006 seems to support this observation.

l The estinates of metallic component failure probabilities were based on:

a) accidental pressurization peak pressure limited to the initial RCS pres-sure of 2250 psia.

b) a probability of f ailure at the yield strength of the material to be 0.01 and the probability of failure at the ultimate strength of the material to be 0.99.

c) the characterization of the overpressurization event as a quasi static process.

d) the statenent that at 2250 psia, the stresses in the limiting RHR piping are only approaching yield stresses and the heat exchanger tube and other riechanic al co-conents are at a small fraction of their respective yield stresses.

a s 2-16 The characterization of the overpressurization event as a quasi static process with a limiting peak pressure equal to the initial RCS pressure of 2250 is based on IDCOR evaluations which have not been reviewed by BNL. The assignment of it and 99t failure probabilities to the yield and ultimate strengths of the material respectively is acceptable since a failure at yield is considered unlikely while a failure at ultimate is considered very likely.

The statement concerning the safety margins inherent in the RHR piping and retallic components and the basis for their calculation, however, must be substantiated. Further, the influence of life or time dependent effects on these safety margins must also be considered. Of particular concern in this regard is the capacity of the potentially corrosion degraded or embrittled heat exchanger tubes to withstand any dynamic loads associated with the over-pressurization event.

Tne following request for further information has been made. "In the description of RHR pressure Boundary Failure Modes, it is stated that the naximun value of stresses due to pressurization to 2250 psia in the limiting s

RHR piping are approaching the yield stress and the stresses in the other metallic co ponents are at a small fraction of their respective yield stresses. Describe the analyses conducted to support this conclusion and provide a sunnary of the pertinent results. In addition, clarify whether the pressure loading has been applied as a dynamic pulse coupled with corrosion degradation effects (such as heat exchanger tube embrittlement). If these effects have been considered, describe the analyses and the dynamic loads. If not, provide the basis for not considering these effects."

Tne applicant's response to this request will be reviewed and the results of our evaluation discussed in a future supplement to this report.

2.1.6 Event Tree Quantification This section summarizes the effect of observations made by BNL in previous sections to the event tree quantification.

Fron the above observations, it is obvious that the main problen in the quantification of various ISL scenarios are related to the determination of 1

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2-17 the initiator frequencies. The other observations and questions expose mainly the overall uncertainty of the real frequencies of these accident scenarios, j The effect of the change in the initiator frequencies to the plant damage states can be demonstrated if the new initiator frequencies, VI and VS, given in Sections 2.1.3.2 and 2.1.3.3 are propagated through the corresponding event trees.

Table 2.5 presents the results of the RNL requantification. The table and its notation is essentially the same as Table' 3-14 of the Seabrook EPZ Study. 'For convenience, in Table 2.5, the meaning of some plant damage states are repeated.

From the table, the new value of the total core damage contribution due te ISL can be deternined (the sum of PDS states BC through IFV). This is:

CDI st = 1.37X10-6 event / reactor year.

l The value obtained is much higher than the updated value (see Section 2.1.1) o' the Seabrook EPZ Study. It is much closer to the result of an earlier assessnent given in the SSPSA, which is, CDISL = 1.8x10-6 event / reactor year.

2.2 Accider.ts During Shutdown and Refueling Conditions The Seabrook Energency Planning Study 2 concentrated on accidents that would occur during power operation, and did not assess the risk during non-power operation. Table 2.6 defines 6 modes of plant operations as were defined in standard technical specifications. As far as early releases are concerned, there are some potentially significant contributors from operation in nodes 4, 5, and 6. Typically, technical sp'ecifications do not address the status of containnent isolation in mode 5, and require isolation in mode 6 only during periods of fuel handling. Consequently, it is possible to have a core nelt accident with the containment wide oper.

4 9

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2-18 NSAC-8:e is the only study that was performed to assess the core damage frequency due to accidents during non-power operation at PWR. It is an inno-vative and detailed study for the Zion plant, using the plant-specific proce-dures and experience. Three types of initiating events were considered: loss-of cooling, low temperature overpressurization, and loss of coglant. NSAC-84 results show that the dominant core damage sequences are due to loss of RHR systen and hunan errors. The contribution of LOCA to core damage frequency during shutdown and refueling is approximately 2x10 8/ year. The contribution of low temperature overpressurization is assessed to be less than 10 IO/ year.

The total core damage frequency during shutdown or refueling was assessed to be 1.8x10-5/ year which is comparable to the frequency of core damage during power operations, i.e., 5.7x10 5/ year. It was stated in the executive sunnary of NSAC-84 that "with the uncertainties involved, the risk of fuel damage during some period of a shutdown may be as great as the risk at power."

BY. perforned a limited review of NSAC-84 Based on the limited informa-tion docunected in NSAO-84 and limited information for Seabrook, BNL believes that the dominant sequences identified in NSAC-84 may also apply to Seabrook, while the frequencies of the accident sequences need to be reassessed. The NSAC-P4 analysis of low tenperature overpressurization may be too optinistic.

Events such as those of Turkey Point-48 indicate that the frequency with which a rapid pressurization occurs with the RHR system isolated and the p0RVs unavailable is higner than 10-3 per year. The operators have only a few minutes to respond to the event before the pressure reaches the setpoint for i the saftty valves. Therefore, the human crror probability is not going to be very small. If the operators f ail to terminate the overpressurization, the primary systen pressure will reach the setpoint for the safety valves. At this pressure, the vessel rupture probability may be of the order of 10-3 10-11 The following sections discuss the possible initiating events in nore detail. Operational experiences and causes of failures are provided for each type of initiating events. Some scenarios that may lead to core damage are provided based on reports in the related area. Related safety issues are also provided.

n , - m.-, . , . _ . . _ - . - .

D e 2-19 2.2.1 Loss of Decay Heat Removal The RHR systen is designed to remove decay heat from the primary coolant syiten during modes 4, 5, and 6. Most of the time when the system is oper-ating, only one train is actually running; the other train is either on stand-by or unavailable due to test or maintenance. If the operating train fails to continue running, the standby train will not start automatically. Therefore, loss of the operating train leads to loss of the system and operator actions will be required to restore it. The Office of Analysis and Evaluation of Operational Data (AE00) identified and analyzed 130 loss-of-DHR events at PWRs during approximately 500 reactor years of operations.12 According to this experience base, the f requency of loss-of-RHR is estimated to be 0.25 per reactor year. Table 2.7 lists the categories of the 130 events. it can be seen that automatic closure of suction valves and inadequate RCS inventory are the two dominant causes of loss of DHR. Automatic closures of suction valves were caused by spurious high pressure signals, loss of instrument bus, and huran errors in calibration of pressure transmitters. Inadequate RCS inven-t ory was caused by hunan errors and inadequate vessel indications during drained-down operations. It was estimated 12 that approximately two-thirds of the events were human error related.

Upon loss of DHR, operators may be able to restore the failed train by reo;;ening the spuriously closed suction valve, or starting the standby train.

Alternative rethods for decay heat renoval include use of steam generators and use of charging purpt. Typically several hours are available before core uncovery occurs. Ineref ore, the most important thing is that the operators must be able to recognize the loss of DHR. In the 130 loss-of-DHR events identified in the AE00 study, the operators responded in a timely fashion, such that no serious damage resulted. However, the duration of loss of DHR in sore cases exceeded one hour. If a loss of DHR occurs 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after a reactor trip with the plant in a partially drained Condition, the onset of core uncovery may be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> af ter the loss-ofDHR.12 ine dominant core damage sequences in NSAC-84 represent scenarios in w':i c h decay heat removal is lost and the operators fail to determine that action to restore cooling is required. For example, the accident sequence

e .

2-20 with the highest frequency is a sequence in which the RHR suction valve is inadvertently closed, the operator fails to trip the RHR pumps, and also f ails to determine that action to restore cooling is required. Its frequency is estimated to be 4.3x10 8 per reactos year. Such scenarios can be postulated for Seabrook. However, quantitative assessment of the accident scenarios must take into consideration the plant-specific information. For example, operator performance is strongly affected by the indications or alarms available in the control roon. Zion does not have any alarm in the control room for inadver-tent closure of the RHR suction valves, while the Seabrook control room has an audible alam on the video alarm system if the RHR pump is running with the suction valve closed. Another difference between Zion and Seabrook is that Zion has a single drop line and Seabrook has double drop lines. This is not expected to be a significant difference because the auto closure logic at Sea-brook will isolate both suction lines when a spurious signal is generated, i.e., a single pressure transnitter provides input to the interlock logic of the two inr.er isolation valves, and a separate pressure transmitter provides input to the interlock of both outer isolation valves. BNL performed a LER searen for less-of-042 events due to spurious closure of suction valves at plants with decle drop lines. Seven events were found in approximately 40 reacter years. This indicates that the frequency of spurious closure of suction valves at plants with double drop lines is not lower than the average frequency for all plants.

In response to the NPC request for additional information, the Public Service of New Ha: pshire (pSNH) provided a shutdown risk analydis for Sea-brook 33 utilizing the results of NSAC-84 Two differences between Seabrook and Zion were accounted for, viz., the number of hot leg suction lines, and the support syste": interf aces with the RHR systen. In the analysis, credit for the additional suction line at Seabrook was.taken, based on the statement "For spurious valve closure to cause a loss of RHR cooling at Seabrook station, it is necessary to postulate either a common cause event involving one valve in each suction path, or a coincidence of a single valve closure and naintenance being perforned o:. the other RHR train." This reduced the fre-quency of loss of RHR by a nultiplicative factor of 0.145 and the core danage frecuency by approximately a factor of 2. From the information available to B 'R , it is not cleaa that this reduction is justified. It is true that a

2-21 single train of the RHR systen is adequate for decay heat removal. However, the standby train is not normally operating and will not start automatically when the operating train becomes unavailable. The analysis in reference 13 assunes perfect automatic start signal for the standby train or perfect opera-tor response to the loss of the operating train; however, it has previously been shown that operator error is important in' these sequences, and neglect of it here is inappropriate.

The analysis of the support systens in reference 13 has not been reviewed by BNL. As was stated in the analysis, the differences in the support systen interfaces with the RHR system are unfavorable for Seabrook.

Two issues are related to the availability of RHR system, i.e., unre-solved safety issues A-45 and generic issue 99. A-45 addresses the adequacy of decay heat removal systems in existing light water reactor nuclear power plants. Generic issue 99 addresses the RHR suction line interlocks on PWRs.

B'il is currently involved in a project to investigate methods to inprove the reliatility of RHR systems during shutdown or ref ueling. The results of the pecject will be used towards resolution of generic issue 99.

2.2.2 Low Te perature Overpressurization Low ter;erature overpressurization may occur during shutdown as a result-cf unanticipated addition of nass to the reactor coolant systen, for example, inadvertent actuation of safety injection pumps, or imbalance of letdown and charging flows. Inbalance of letdown and charging flow may be caused by spurious isolation of the RHR systen, thus, violation of the letdown flow, and loss of instrument air that causes the letdown flow control valve to close and the charging line flow control valve to open. To protect the Seabrook plant against such scenarios, a low temperature overpressurization protection system is activated when the primary systen is cooled down after a reactor trip. The systen nonitors the primary system pressure and temperature and actuates a nain control board alarn when the pressure reaches a pre-determined fraction of the allowable pressure, and on a further increase in measured pressure, transmits an actuation signal to the PORVs and the PORV isolation valves.

Alsc, the safety injection punps and one or more of the charging pumps are

s .

2-22 nade inoperable during initial cooldown. In addition to the PORVs the relief valves in the RHR syste9 may be available to relieve the pressure. Each RHR suction line has a relief valve with 900 gpm capacity at 450 psig, and each RHR discharge line has a relief valve with 20 gpm capacity at 600 psig. How-ever, these relief valves may be made ineffective if the RHR suction valves close automatically when the setpoint of 600 psig is reached, as was the case in the Turkey Point-4 events. Actually, the Seabrook Technical Specifications only require either both PORVs or both RHR relief valves to be available.

Two generic issues are related to the subject of low temperature over-pressurization, generic issues 94 and 70. Generic issue 94 considers addi-tional l ow-t erp e ra t u re -o ve rp re s su ri z a t i on protection for light water reat-tors. It has a "high" priority ranking.13 Enclosure 1 to reference 11 is the prioritization evaluation for the issue. It was stated in the evaluation that before 1979 33 events in PWRs were reported where the pressure /tenperature of the rea: tor coolant systen violating Technical Specifications. Af ter 1979, following changes to operating procedures and the implementation of overpres-surizttion ritigation systems, there have been two reported events of over-pressure excuasion events, i.e., Turkey Point-4 events. Based on the opera-tional experience and the use of the Vessel Integrity Simulation Analysis (V:SA) coce,3' the prioritization evaluation estimated that the core danage frequency due to vessel rupture in a low-temperature-overpressurization event at Oconee 3 to be 4.5x10 ' per reactor year. Generic issue 70 considers the reliability of PORV and its block valves. BNL is currently investigating the issue, and a draf t recort of the work is upenning.

2.2.3 Loss of Coolant Accidents

! hSAC-52M reviewed operating experience within 5 calendar years up to the end of 1981, and identified 10 loss of coolant events at PWRs. They were caused by the following causes:

1. Inadvertent nanual initiation of RHRS supplied containment spray.

2. Inadvertent loss of inventory to the containment building sump and/or automatic initiation of recirculation mode of low pressure safety inje: tion, i 3. Inaevertent loss of inventory via the RHR$ relief valves.

. , . _ _ , , _ _ ~ _ . . _ . . , _ -. . _ _ - - - _ _ _

2-23 4

4 Inadvertent loss of inventory via mispositioned crossconnect or drain valves.

5. RHRS valve packing gland removal during plant pressurization, dislodging the valve packing and gland.

6. Gross valve packing leak. . ,

As for the loss-of-cooling initiating event, LOCA 'uring shutdown or refueling requires operator response to terminate the inventory loss and to provide inventory make up. The NSAC-84 analysis for Zion assessed the core i

damage frequency due to a LOCA at shutdown or fueling to be approximately 2x10-6 per reactor year. The doninant scenario is that the operator fails to l close the RHR return valve to the RWST after draining the cavity, on reestab-lishing RHP flow, a LOCA via the RWST vent outside the plant occurs, and the operator fails to respond to it.

2.3 Incute

Steam Generator Tube Rupture (SGTR)

For accidents in which the primary system is at high pressure during core uncovery and nelting, it is possible that large natural circulation flow pat-terns coule develop within the primary system. These flow patterns could in tt.rn heat up regions of the primary system remote from the reactor core. As the primary systen heats-up, it is possible that parts of the pressure boundary could degrade. Of particular concern is the possibility of degrading the steam generator tubes such that the primary system could become open to t

the setonda*y syster, if the secondary system were in turn open to atmo-sprere, then a direct path could exist between the primary systen and the atmosphere, which bypasses containment.

i Tnis is a very important topic for review because it could potentially

) lead to a relatively large early release of radioactivity, and the applicant I

considers it to be very unlikely. Tne topic was not included as part of the work scope for the current BNL review. Howev,er, the topic was reviewed I8 in detail by the fAC staff and is the subject of continuing NRC and industry

research activities.

l i

4

2-24 Scoping studies were performed to assess the impact of induced stean generator tube rupture on risk at Seabrook. First, the frequency of accidents in which the primary system would be at high pressure had to be determined.

The applicant estinated l7 the frequency of high pressure sequences in which a SGTR might have an effect to be 4x10 s per reactor year. This estimate was l

considered reasonable in the NRC review ' and therefore used as the basis for the BNL scoping study.

Given that core meltdown occurs with the primary system at high pressure, the probability that the stean generator tubes will fail had then to be deter-nined. In addition, it is also possible (provided methods are available) for the operatcas to depressurize the primary systen prior to induced f ailure of the SGT. Tne probability of successful depressurization had also to be deter-nined.

Estinating the probabilities of the above events is subject to signifi-cant uncertainty. However, the Severe Accident Risk Reduction Program at SNL attempted to quantify these probabilities by use of expert judgment. The probabilities were developed specifically for the Surry plant and reported in Appendix B of SANDS 6-0119.le The experts concluded that there was a condi-tional probability of 0.8 for successful depressurization of the prinary systen, in addition, they felt that the probability of an induced stean generator tube rupture might be between 0.01 and 0.1 (for both small and large

, tube ruptures) conditional on no depressurization. These estinates are reasonably consistent with an earlier NRC nemorandunII on this subject, whicn suggested a conditional probability of about 0.01 to 0.3 for SGTR given a high pressure core neltdown.

4 It was therefore decided to use 0.2 as the condition-al probability of f ailure to depressurize and a range of 0.01 to 0.3 for SGTR to assess the impact of this phenomenon on risk at Seabrook. The results are sunnarized in Section 2.5.

4 2.4 Containment isolation Failure To be provided.

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2-25 2.5 Su mary In the introduction to this section, several categories of questions were listed; the topics treated here are each particular instances of those general categories. It must be emphasized that there has been no top-down review of the process of establishing the frequency of any release category; the topics addressed here were chosen essentially a priori. For some topics, sensitivity studies have been performed using the applicant's conditional risk indices to show how the dose vs distance and risk profiles might change as a result of the concerns raised in this section. The results are summarized below.

Inter'acinc Syste LOCA The contribution to release categories S1W and 57W from interfacing systen LOCA (as given in reference 1) are reproduced below:

Source Tern Category Plant Danage State Frequency SIW 1FV 4.6x10 9 574 1FPV, 7FPV 3.9x10.e Tne BT. reavantification of the event tree for interfacing systen LOCAs (in Section 2.1) suggested the revised frequencies in Table 2.4 These revised frequencies would imply the following contribution to source term categories $1W and S}W:

Sourca Terr Category Plant Damage State Frequency S1W IFV 1.4x10 7 57W IFPV, 7FPV 1.1x10 6 The inpact of the above changes on the risk estimates in PLG-0432 is given in Table 2.8. The revised frequencies for interfacing system LOCAs have (as expected) no impact on the risk of cancers (refer to Table. 2.8) because this health effect is dominated by the higher frequency S2W category. The irpact of the revised frequencies on early fatality risk is very small if no evacuatior. is assumed. However, as successively larger evacuation distances are assuned the revised frequencies have rather more effect. However, as the no evacuation assumption is already below the proposed safety goals, the f l

1 1

2-26 changes in risk for the one and two mile evacuation assumptions are also ev e within the pn -~~cd safety goals. The above changes have no inpact on the 200 ren dose vs distance curves in PLG-0432 (reproduced in Figure 2.2) because they influence conditional frequencies below 0.001.

Note that in Section 5.2 questions were raised regarding the credit to be taken for pool scrubbing of fission products for the S7W source term category. In the RSS 21, a decontamination factor of 100 was used for subcooled pool scrubbing rather than the factor of 1000 which was used in PLG-0465, to calculate the S7W release fractions. BNL has confirmed that the effect of using a factor of 100 rather than 1000 would not change the results j of the BNL sensitivity study in Table 2.8. However, the statenent made in Section 5.2 that neglecting pool scrubbing for all interfacing LOCA sequences would have no inpact on risk is only true if the frequencies in PLG-0465 are used. If pool scrubbing was neglected and the revised RNL frequencies in Table 2.4 was used, then the risk results would be ef fected. However, BNL does not consider it appropriate to neglect pool scrubbing for that subset of interf acing system LOCA sequences where the break location is under water.

Tnus, BNL censiders the sensitivity results presented in Table 2.8 to be appropriate even when our concerns in Section 5.2 are taken into acocunt.

Accidents Durino Shutdown This topic was not originally addressed in PLG-0465 and , a detailed assessnent of such events is beyond the scope of the current BNL work on this project. However, the applicant was requested to provide information on the risk associated with accidents during shutdown. The results of the

' applicant's assessnent of such accidents were presented in the form of sensitivity studies in reference 13. BNL is not presently in a position to I

assess the f requency of these events for Seabrook becuase, as explained in Section 2.2, there remain fundamental questions regarding the modeling of these scenarios. However, in spite of this, the applicant's results have been included in Table 2.8 and Figures 2.2 and 2.3 for comparison with the BNL sensitivity study results on other topics. It should be noted that the

} applicar. considers their upper bound estimates in Table 2.8 and Figure 2.3 to he very conservative. In particular, in order to assess the inpact of these

2-27 everts, they were included in source term categories derived for accidents from full power, which could lead to predictions of shorter times and larger quantities of fission product release than would be expected from accidents during shutdown.

I As shown by the applicant in Table 2.8 and Figures 2.2 and 2.3, the influence on risk of accidents during shutdown can be significant. It should be noted that this topic is being addressed generically in other ongoing NRC work.

Induced Stea- Generator Tube Rupture In Section 2.3 a sensitivity study was suggested to assess the impact of induced SGTR on risk at Seabrook. The frequency of high pressure sequences taken togetner with the conditional probabilities of failure to depressurize and induced SGTR given in Section 2.3 give the following range of probabili-

, ties of induced SSTR: .

4.0x10 5 x 0.2 x 0.3 = 2.4x10 6 per reactor year 4.0x10-5 x 0.2 x 0.01 = 8.0x10 e per reactor year.

In order to estimate the impact of the above probabilities on risk an appropriate source term category had to be selected. it was decided to allo-cate SGTR events to release category S1W, which represents a large early bypass of the cortainment. It was felt that this was a conservative assumption because significant retention of the fission products in the secondary side could occur and this was not considered when calculating the SIW release fractions. The impact of adding the above frequencies to source term category 51W is illustrated in Table 2.8 and Figure 2.4 The impact of the above changes relative to the risk of cancers given in PLG-0465 is again relatively small even when the higher probability of SGTR is

),

assumed. However, the impact on early fatality risk of the highest proba-bility of SGTR has rather more impact. The risk of early fatalities increases from 2.7(-3) to 4.5(-3) per reactor year assuming no evacuation and remains relatively higa even with a two mile evacuation. The risk of early f atalities i

2-28 assuming no evacuation and the high SGTR probability is close to and nay a (LC-- . . -

exceed the md safety goals. It* should be noted that the risk estimates in Table 2.8 are sensitivity studies and that the NRC staff believes!' that the actual probability of 'a SGTR is closer to the lower estimate. However, given the uncertainty associated with predicting such events it is prudent to indicate the impact on risk of a range of assumptions regarding SGTR.

The inpact on the 200 ren dose vs distance curves in PLG-0465 is shown t in Figure 2.2. Only the more conservative probability for SGTR has an impact on the 203 ren dose vs distance curves and then only on conditional frequen-cies below 0.01. The statement in PLG-0432 that "the risk of radiological exposures foe 200-re9 whole body dose with no immediate protective actions is less at I nile than the corresponding NUREG-0396 results at 10 niles" is not affected by this particular sensitivity study related to the potential for SGTR.

Centainrent Isolation Failure (to be provided) 2.6 Refere9:e_s

1. PLG-0465.

2. Seabrook Station Risk Managenent and Energency Planning Study, PLG-0432, December 1985. ,

3. SSPSA.

4 MAAP code.

t 5. NUREG/CR-1363.

6. Tean Enhanced Evaluation Method (TEEM) - Checklists and Instructions, Infornal Report BNL-38585, September 1986.

7. Handbook of Hunan Reliability Analysis With Emphasis on Nuclear Power Plant Applications, Final Report NUREG/CR-1278, August 1983.

8. Zic* Nuclear Plant Residual Heat Removal pRA, NSAC-84. July 1985.

9. "Overpressuri:ation of Reactor Coolant System," IE Information Notice hc. 82-17, U.S. NRC, June 11, 1982.

1 l

2-29

10. T. J. Burus et al., " Pressurized Thermal Shock Evaluation of the Oconee-1 huclear Power Plant," NUREG/CR-3770, Draft. April 1984

11. Meno f rom H. R. Denton, Director, Office of NRR, to R. M. Bernero, Direc-tor, Division of Systen Integration, on the Subject of Schedule for Resolving and Completing Generic Issue No. 94, July 23,1985.

12. H. Ornstein, " Decay Heat Renoval Problems at U.S. Pressurized Water Reat-tors," Office for Analysis and Evaluation of Operational Data, d.S. NRC, Decenber 1985.

13. PSNH letter (SBN-1225), dated October 31, 1986, " Response to Request for Additional Information (RAls)," J. DeVincentis to S. M. Long 14 NUREG/CR-338", " VISA--A Conputer Code for Predicting the Probability of Rea:ter Vessel Failure," U.S. NRC, September 1983.

15. Residual Heat Renoval Experience Review and Safety Analysis, NSAC-52, Janua ry 1983.

l

16. Warren Lyon's neno on SGTR.

17 PSN" Letter (SSN-1237), dated Novenber 21, 1986, " Emergency Planning l Sensitivity Study," J. DeVincentis to S. M. Long.

18. SAN 796 n!!9.

19. h?C Menorandun, dated February 14, 1985, " Steam Generator Tube Response during Severe Accidents," B. W. Sheron to D. B. Liaw.

20. EGS-NT A 72F.7.

21. RSS.

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Figure 2.2 Comparison of 200 rem and 50 rem dose versus distance curves with contributions from shutdown events.

(Calculations performed by applicant)

t , , , , ,

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- - - = = = PLG-0465 o -

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~

curves for conservative assumption of no credit for operator recovery of open equipment hatch. (Calcula-tions performed by applicant) l l

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1

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1 Figure 2.4 Comparison of BNL sensitivity study related to SGTR with PLG-0465 and NUREG-0396 (200 rem plots with no imediate protective actions).

l

e T.shle 2.1 Sumanry of Oparnting Evants, rmarqanc y Cora roolinq e,ystam, isniat ion Char k Valves, tenhage rallure *de Nieber of Chack istiantad Dalaranea (CCS Valvas taak pata (NIT o Plant Data Systa Fvant Dascription Paported (qrmt pamarks vil.a.1% r% I l s a<ias 5/77 Arc taak nqa Into $1 tank. Tha internals of a chwk valva on the outlet of an Si tank 1 y5 tb te 1 was Incorrm f ly assmblad.

vll.4.25 reign 12/72 Arc tnahaga into Si tank. A small pin e of wald sinq had lodqad undar the saal of the vanhan 1 y5 thte 1 ou t l et c ho- k va l va a l low i nq h.vi k l aa k aq a, Ollution: 1700 p m (limit is 1770 pm).

vit.A.32 Tur k ay 5/73 eiPl One of the thran chnck valves in tha $1 linns davaloped a tankage of 1/3 qpm. 3 g.33 Point Two other c hm k valvas shn=no only slight l aak ag e. Fallure of soft seats.

Vll.A.63 Glnna 9/74 ACC Laaknqa of a chack valve causad baron dilution in ACC. "A" (f rc= 2250 ppm to 1 y20 Note 1 1617 pgn ) .

Vill.4.85 Surry 1 8/75 ACC Chack valve did not seat. ACC ("lC") leval increased. Leshnge rate: a4 gpm. 1 y10 Note 1 vil.A.176 Zlon 2 10/75 ACC Wrong sire ansket Installed In the check vntve for ACC. "A". Leak rate: =.25 gpa. I t.25 Note 1 Vll.A.105 Robinson 2 1/76 ACC Accipulator ("A") Inteakage through lesking outlet check valve. 1 y20 Note 1 v.A.177 Zion 1 6/76 ACC Inteakage to ACC. "10" from RCS. 1 y20 Note 1 Vll.A.114 Surry 1 7/76 ACC Two check valves in serlas (1-St-128,130) leaked causing boron dilution in 2 In series y10 Note 1 ACC. "0".

Yll.A.120 Surry 2 8/76 ACC lbron dilution (f rom 1950 pone to 1893) In St ACC. "C" caused by leeking check 2 in series y10 Notes 1 valves (7-51-145,14 7). and 2 Vll.A.225 Nittstone 2 4/77 ACC Inteakage of RC through outlet check valves to Si tank Low boron "4". 2 In series y20 Note 2 concentration. Five occurrances In 1977 VII.A.175 San 5/78 LPI Tilting disk check valve (first valva inside conteln=ent) f ailed to close with Onofra 1 gravlty Installed In a var t ical rathar than a hortionfal piraline. 1 y5 V i ll. A.182 Calvert 9/78 ACC Outlet chev-k valves for Si tanks 21R and ??R leaked. Roron concentration reduc- 2 y10 Note 1 Cliffs 2 tion from 1724 and 1731 pra to 1652 and 1594 ppm In one month pariod. y10

o Talile 7.1 inn t i nn e.1 Mebar o f Chec k Istfantad Pn f erane n F CCS valvas leak Data

( PF[ f) I'lant Data Sys t m Fvant DascrOptinn Raported (qrm) Ramarks v i l. A.2s 2 Crystal 7/80 ACC Ch ac k valvo Cf v-79 to enra f innd tank fallad, ihn isolation valve to the N 7

1 100<y Notes 1 plvar 5 systm was opnn for Ny mlving, s#00 qallnn liquid entered the Ny syst m and i

(200 and 2

=20 qalinns was reinasnf, ihn corrnspondinq v t ivi ty relaasno est imated as 1.07

• Cl.

Vll.A.273 Davis 10/80 ACC RHR systm isniat ion chnc k valve Cr-30 leakad bac k ewe nssively, valve 1 50< p100 pete 1 lf Info. flassa 1 disk and arm had snparatad f r<n the valve body. Onits and inc k inq machanism Pet le n were missing. Core flood tank overpressur lind.

80-41 vil.A.291 Surry 2 1/81 ACC Accumulefor ("C") boron dilutad. Chm k valve (1-5 8-144) leakad. Flushing syst e 1 y10 Note 1 Improperly set up, resulting in charging syste prassure to avist on the downstream side of the check valva.

Vll.A.301 Palisadas 3/81 ACC Leakage of RC Into the St. tank (T-823). 1 y5 Notes 1 and 2 vil.A.306 W Guire 1 4/81 ACC kcumulator "A" outlet chnck valves IW159 and IW160 were leaking. RCS pressuret 2 In y10 Note 1

, 1800 psiq. Mc. pressure: 425 psig. Water level above alarm setpoint, series Vll.A.307 M-Guire 1 4/81 ACC Similar events with kc s. "C" and "0". 2x2 in y(10 Notes 1 serlos y10 and 2 v il. A. 54 3 Point 10/81 LPI RCS/LPI isolation check valve Il-853C) f eeks la eucess of acceptance criterle 1 y<10 Dny h 1 l >6 gon ) .

Vll.A.3ne Calvart 7/82 ACC &c. outlet check valve at tinit 1 f eaked due to deterioration of the disk seeling Note 1 1

E200 Clites o-ring. The o-ring matarint has baan changed on all Check valves of Unit 1 and 2 1&2 1/2 51-215, 225, 235, and 245 V ll. A.405 .Surry 2 9/82 ACC Mc. outlet chmk valve 12-51-1849 laaked RCS water into tank "C" during a pipe 1 y20 petes 1 flush resulting In low horon concentration, and 2 vil.A.396 Pallsadas 9-12/ ACC Minor lenhage into $1 tank (compoundad by f evel Indication f ailure) vie check 1 y<5 Notes 1 82 valve laaknqas.

and 2

e Tafile 7.1 Cont inuni thenbar of Chack fstimatad Da t ar av e FCCS SNPr #1 Valvas leak Rate PInnt Date Systm Fvant Dawrlption Reported (g ewa l Ramarks Vll. A.40 7 St. Cu l r e 1 5/83 ACC RCS watar in laak nqe t hrowth ou t lat c hev' k valves IS170 and 1&l71, resulting  ? In series 20< p %0 tbte 1 In low boron com entrat ion i n CI A af t".

Vll.A.437 rartny 2 9/83 HPl $1 check valve to loop 3 cold inq was emessively Insking, incomplete contac t 1 50< p 100 botunan the valve disk and saat.

tFR R4-001 (tonen 1 3/84 ACC Ace tenul a tor I"A") Infanknqe through laak ing valves. Administrat ive def ic iency, no manaqmant control over a known probleen (sinca 8/A33 2 in series 75 Note 1 V.F.0043 Palisadas 7/A4 ACC Accumulator intaakage through laaking chec k valves CF-3146 and CK-3116  ? y(5 Notas 1 tfR R4-012 and ?

Vll.A.457 St. L uc ie 12/84 ACC Inteakage to Si tank. Sant piste c<thed, valve saat compensating joint ball 1 20< p50 Note 1 2 galled.

Vll.A.456 Calvert 1/85 ACC Intenhnge to saf ety injaction tanks through check valve, o-ring material 2 y5 Note 1 C91895 degradat ion (Uni t 1 = 1.6 gem, Un i t 2 = 2 7.2 gem ) . 20<y50 1&2 '

t, Vll.A.457 9tGuire 1 4/85 ACC Lov accumulator boron concentration. 1 y5 Note 1 LER 85-007 Pallsadas 6/85 ACC Inleakage from the RCS via a check valvo. Low level boron concentration. 1 y5 Note 1 V ll.4.4 74 Palisades 11/85 ACC Accumulator ($1T-8201 Intenhage from RCS through a checit valve, CK-3116 1 y5 Note 3 Boron dilution.

Note 1: Estimated leak rate is the resultant one through two Check valves in series.

Note 2: Not listed in Table 3.R of PLG-04 32.

Hofe 3: The Palisades unit has a chronic accunulator Inlenkaqn prohlten.

O e

T.s fil e 2.2 Sir-ary of Oparat tnq Fvants, fearqancy rnra ennlinq Systam. Isolatinn Chnc h valvas. "Falture to Close" Fallure **)de N eher of Chack onf aranc a FCCS Valvas frJrF #i Plant Date Systm Fvant Description Raportad Rearks Vll.A.770 Saquoyah 1 9/80 601 S i c hac h valva 6kMs was f ound to ha stic k opan, it was caused by 1 Note 1 Inter f arenca hatwnan tha disk nut lockwIra tack wald and the valve body.

Vll.A.7ns Salm 1 17/R0 6PI $1 check valva f allad to closa during a test, it Is an Ontarf ace betwaen RCS 1 teta 1 hot inq and $1 pumps, valva was f ound to be inc hed open dua to horon solldl f ica-tion during the last rnf uallnq.

vil.A.704 (tonaa 1 7/81 LPI Reac tor vassal LPI loop "f1* Isolation valva ICCF-173 lankad aw:essively during 1 Pete 1 LOCA lank test. The valva disk had hnenma f roran at tha pivot In a coc k ed position. Ilulldup of deposit in the gap hatwaan the hInga and disc linob caused the fraaring.

vil.A.307 Oconen 3 3/81 LPI Similar to event at twit 1 tvalve involved is 3 CF-131 1 Note 1 vil.A.310 stGuire 1 5/81 ACC teak test dmaqad acc. chack valves - seat type chanced. 2 Note 1 vil.A.311 etGuire 1 S/81 ACC Acc. check valves f alled. 2 Note 1

.vil.A.315 Point 7/81 LPI RCS/LPI Isolation check valves 1-853 C and D were found to be stuck in the full 2 Note 1

, Danc h 1 open position. High leakage rate.

vil.A.397 ANO-2 10/R2 tel $1 Isolation check valves 2 SI-13C and 2 $l-139 stuck In the open position during 2 sete 1 test requested by IE Notleo 81-30. Disk stud protruded above nut, disk else11gned.

Note 1: PMt listed in Table 3.9 of PtG-0437

. _ _ _ . ~

Table 2.3 Accumulator Check Valve Exposure Data Start of Number of Total Number of Commercial Number of Accumulator Check Valve Plant Nare Operation Years Check Valves (105 Hours)

Arkanasas Nuclear One 1 December 1974 11.08 4 3.882 Crystal River 3 March 1977 8.83 4 3. 09 /.

1 Davis-Besse ! November 1977 8.16 4 2.859 Oconee 1 July 1973 12.50 4 4.380 Oconee 2 March 1974 11.83 4 4.145 Oconee 3 December 1974 11.08 4 3.882 Rancho Seco April 1975 10.75 4 3.767 Three Mile Island 1 September 1974 11.33 4 3.970 Three Mile Island 2 December 1978 7.08 4 2.481 Arkansas Nuclear one 2 March 1980 5.83 8 4.086 .

Calvert Cliffs 1 May 1975 10.67 8 7.478 Calvert Cliffs 2 April 1977 8.75 8 6.132 ,

Fort Calhoun September 1973 12.33 8 8.641 Millstone 2 December 1975 10.08 8 7.064 Maine Yankee Dece=ber 1972 13.08 6 6.875 Palisades Dece=ber 1971 14.08 8 9.867 St. Lucie 1 December 1976 7.08 8 6.363 Beaver Valley 1 April 1977 8.75 6 4.599 D. C. Cook 1 August 1975 10.42 8 7.302 D. C. Cook 2 July 1978 7.50 8 5.256 Indian Point 2 July 1974 11.50 8 8.059 Indian Point 3 August 1976 9.42 8 6.602 Joseph M. Farley 1 Dececher 1977 8.08 6 4.2/7 I

Kewaunee June 1974 11.58 4 4.058 North Anna 1 June 1978 7.58 6 3.964

Prairie Island 1 December 1973 12.08 4 4.233 Prairie Island 2 December 1974 11.08 4 2 3.882 Point Beach 1 December 1970 15.08 4 5.264 Point Beach 2 October 1972 13.25 4 -

4.643 R. E. Cinna 1 March 1970 15.83 4 5.5a7 H. E. Robinson 2 March 1971 14.83 6 7.795 Salem 1 June 1977 8.50 8 Surry 1 5.957 December 1972 13.08 6 6.875 Sorry 2 May 1973 12.67 6 Trojan 6.659 May 1976 9.67 8 6.777 Turkey Point 3 December 1972 13.08 6 6.875 Turkey Point 4 September 1973 12.33 6 ~6.481 Yankee Rowe June 1971 14.50 2 Zion 1 2.540 1

December 1973 12.08 8 8.466 Zion 2 September 1974 11.33 8 7.940 McGuire 1 December 1981 4.08 8 Sequoyah 1 2.859 July 1981 4.50 10 3.912 Sequoyah 2 June 1982 3.58 10 3.136 TOTA'-

2.369(2)

Table 2.4 Statistical Data on Leakage Events of Check Valves to Accumulators Leak Rate (gpm) Frequency of Frequency of Nue.ber of Events Occurrence (per hour) Exceedance 10 4*64(-7) 1.48(-6) 9 20 9 3.80(-7) 1.01(-6) 50 3 3.80(-7) 6.33(-7) 100 1.27(-7) 2.53(-7) 1 200 4.22(-8) 1.27(-7) 2 8.44(-8) 3.44(-8)

I

\

1 l

, 2 40 Table 2.5 ISL Results initially Assigned Plant Damage States Frequency Contribution From Plant Total Damage State VI. VS Frequency LOCA 1.96-4 1.44-4 3.4-4 DLOC 1.2-5 0 1.2-5 DILOC 9.8-8 7.7-6 7.8-6 8C 2.1-8 0 2.1-8 7D 1.5-7 0 1.5-7 7FPV 7.4-8 1.7-7 2.4-7 IFPV 1.8-8 8.0-7 8.2-7 l IFV 8.4-8 5.9-8 1.4-7 Totals 2.1-4 1.5-4 3.6-4 Note:

LOCA: denotes a PDS, which contains those sequences in which the RC leakage in both ISL pathways analyzed exceeds 150 gpm, but does not exceed the RHR system relief valve capacity.

Tne sequences are essentially medium LOCAs.

DLOC: denotes a PDS, which contains sequences in which the ISL is terminated.

DILOCA: denotes a PDS, in which coolant makeup is being supplied to the core, but the ISL has not been terminated.

The other plant damage states are involving containment bypassing ISLs and core danage.

-,;s n

l 2-41 Table 2.6 Plant Operational Modes

• Average Reactivity  % of Rated Coolant Operational Mode Condition, Keff Thermal Power ** Temperature

1. POWER OPERATION > 0.99 > 5%

> (TDHR) F

2. STARTUP > 0.99 < 5%

> (TDHR ) F

3. HOT STANDBY < 0.99 0

> (TDHR) F

4. HOT. SHUTDOWN < 0.99 0 (TDHR) F>T,yg>200*F

5. COLD SHUTDOWN < 0.99 0 < 200*F

6. REFUELING *** < 0.95 0 < 140*F-

, TDuo = temperature at which the DHR system is initiated (generally 280'F -

350'F

'As defined in B&W, CE, and W standard technical specifications.

Note nany plants do not use standard technical specifications.

** Excluding decay heat.

• • • Fuel in the reactor vessel with the vessel head closure bolts less than -

fully tensioned or with the head removed.

2-42 Table 2.7 Categories of 130 Reported Total DHR System Failures When Required to Operate (Loss of Function) at U.S.

PWRs 1976-1983 No. of Events (% of Events)

Autenation closure of suction / 37 (28.5) isolation valves Loss of inventory Inadequate RCS inventory resulting 26 (20.0) in loss of DHR pump suction loss of RCS inventory through DHR 10 (7.7) systen necessitating shutdown of DHR systen Conconent Failures Shutdown or failure of DHR pump 21 (16.2)

Inability to open suction / 8 (6.1) isolation valve Others 28 (21.5)

Total 130 (100.0)

2-43 65I Table 2.8 Inpact of BNL Sensitivity Studies on PLG-0438 Risk Estimates Absolute Risk Per Reactor Year Early Fatalities Total Cancers Sensitivity One Two One Two No Mile Mile No. Mile Mile Study Evac. Evac. Evac. Evac. Evac. Evac.

Original 2.7(-3)* 3.6(-4) 2.4(-5) 1.5(-2) 1.4(-2) 9.2(-3)

PLG-043f 65:

Results Revised 2.8(-3) 4.6(-4) 1.2(-4) 1.5(-2) 1.4(-2) 9.3(-3)

Frequency for Inter-facing Systen LOCAs Stear **

4.5(-3) 2.2(-3) 1.8(-3) 1.7(-2) 1.6(-2) 1.1(-2)

Generator Tube 2.8(-3) 4.2(-4) 8.4(-4) 1.5(-2) 1.4(-2) 9.3(-3)

Ruptured Accident **

6.2(-3) 2.9(-3) 3.0(-4) 2.1(-2) 2.0(-2) 1.4(-2) fron Snut- ***

2.7(-3) 3.8(-4) 2.8(-5) 1.5(-2) 1.4(-2) 9.2(-3) down****

• 2.7(-3) = 2.7x10-3

• • Pessinistic assumptions.

• • • 0ptinistic assunptions.

• • • • Calculated by tne Applicant, not confirmed by BNL.

3-1

3. EVALUATION OF CONTAINMENT BEHAVIOR 3.1. Capacity at General Yield Seabrook Containment Building The Seabrook Station containment building (See Fig. 3.1) is a reinforced concrete structure consisting of a basemat, a cylindrical wall and a hemi-spherical dome.1 Tne basemat is essentially a 10' thick circular slab which supports the cylinder and other internal structures. The cylinder has an internal diameter of 140', a height of 149' and a minimum wall thickness of 4'-6". The dome internal diameter is 69'-11 7/8" and the minimum wall thick-ness is 3'-6 7/8". In addition, the containment has a mild steel liner on the inside. The liner thickness is 1/4" at the base, 3/8" in the cylindrical wall, and 1/2" in the dome.

Tne containnent is reinforced with ASTM ASIS grade 60 reinforcing bars of various sizes, mainly #18, #14 and #11. The specified yield strength for the reinforcing bars is 60 ksi. Median yield stress and lognormal standard devia-tion obtained from the results of tensile tests are shown in Table 3.1.2 por the ultimate strength of #18 reinforcing bars a mean value of 109 ksi and a Cov of 0.025 has been reported.2 The liner steel conforms with ASME SA516 grade 60 for which the specified yield strength is 32 ksi. The mean yield stress was found to be 45.4 ksi with a CoV of 0.042.2 At 271'F a mean yield stress of 43.5 ksi and a Cov of 0.065 are reported.2 The mean ultimate strength at 271*F was estimated to be 59 ksi and with a CoV of 0.09.2 ,

i l

l 1

3-2 1

The primary membrane reinforcement in the cylindrical wall is divided into two equal groups placed near the inside and outside faces of the contain-ment wall. Each group consists of two layers of hoop bars and one layer of meridional bars as shown in Figure 3.2. Since the cylinder basemat intersec-tion is subjected to high bending moments and shear forces, secondary meri-dional reir.forcement is placed in this region (See Figure 3.2). In addition, two layers of seismic rebars inclined 45' to the vertical axis are placed near the c ',ide surface of the cylinder wall. Shear ties are also placed in the cylir near the cylinder-basemat intersection (See Figure 3.2). Major rein-forcerent details for the containment wall are summarized in Table 3.2.

The dome reinforcement follows the cylinder reinforcement until 9.4*

above the springline. Between 9.4' and 79.2' the hoop reinforcement is recuted to one 718 bar near each face. Above 79.2* tr. hoop reinforcement is terminated and tne reinforcement pattern is orthogonal. The meridional cylin-der reinforcement is continued to 60' above the springline with an increase in its density as the elevation increases. Above 60' every alternate meridional reDar is terminated, ano they are bent such that the reinforcement pattern near the done apex is orthogonal. Details of the dome reinforcement are shown in Table 3.3.

Concrete with two design strengths were used in the Seabrook containment building. A 4,00') psi design strength concrete was used for the basemat, for the cylinder near the intersection with the basemat, and for both the cylinder and dome near the dome-cylinder intersection. In the cylinder and dome portions where primary membrane behavior is expected a 3,000 psi design strength concrete was used.

The median and lognormal standard deviation for the 4,000 psi and 3,000 l psi design strength have been reported for 28-day old cylinders and for aged concrete.2 Tnese quantities are given in Table 3.4 obtained from Ref, 2, 1

I

3-3 Seabrook Containment Model A finite element medel of the Seabrook containment was developed to be used with the computer code NFAP.3 The model is shown in Figure 3.3 and is based on an axisymmetric idealization of the geometry, which is considered a 4

good approximation for a structural failure evaluation under axisymmetric pressure loads. The containment finite element model consists of 408 eight-noded isoparametric elements and 1401 nodes. A set of nonlinear spring elements with a bilinear stress-strain law are used to model reinforcing details such as shear ties. In addition, a set of nonlinear spring elements with high compressive stiffness and zero tensile stiffness are placed under the basemat to model the foundation conditions.

I Throughout the cylinder wall and dome the model has 8 layers of eight-noded elements across the wall thickness, as shown in Figure 3.3. Six layers of elements were-used through the thickness of most of the basemat. The element layers and its properties were chosen to represent separately the liner, the plain concrete, and the reinforced concrete with hoop, meridional and diagonal rebars. Spacing and sizes of the layers have been chosen in order to model the actual rebar placements as close as possible. This is particularly pertinent at the cylinder-basemat intersection where high bending moments and shear forces will develop. In addition to these criteria, the modeling requirements commonly used with finite element analysis were also taken into consideration.

The inelastic behavior of the plain concrete is' described by the Chen and Chen elastic-plastic-fracture model.3 Material properties for this model were estimated from the aged (as built) concrete properties and are shown in Table 3.5. Post-cracking behavior of the concrete was modeled using a normal stiff-ness reduction factor a of 10 ", and a shear stiffness reduction factor 8=0.5/(cl/cto), where el is the principal normal strain normal to the

3-4 crack and cto the tensile strain at crack initiation.3 The shear stiffness reduction factor is limited to a value not less than 0.10 to account for the cummulative effect of interface shear transfer and dowel action. The normal stiffness reduction f actor a reduces the normal stress diagonal element in the stress-strain matrix, and 8 reduces the shear stress diagonal element in the stress-strain matrix. The tension stiffening effect was modeled with a f actor -0.1, which multiplies the concrete Young's modulus.

The elastoplastic behavior of the reinforcing bars and liner steel was moceled by a bilinear stress-strain curve and a Von Mises plasticity model with isotropic hardening. Since the #18 reinforcement bars provide most of the reinforcement, the mean material properties for these bars were used for all reinforcing bars. For the liner, the mean properties at 271'F were used.

The plain concrete properties used are shown in Table 3.5.

Loads included in the analysis are the dead weight of the containment and internal pressure. Tne dead weight is applied to the containment in the first load step at the beginning of the analysis, while the pressure load is applied to the containment in small increments (5.0 psig) in order to detect the initiation of nonlinear concrete behavior and concrete cracking. Once the concrete cracks its stif fness in the direction normal to the crack plane is redaced by the factor a defined above, and the released stresses are redistributed to the reinforcing steel. As the pressure load is increased the next nonlinearity is the yielding of the liner steel. At this internal pressure the containment is cracked in both the cylinder and dome, and some flexural cracking has been initiated in the cylinder-basemat intersection region. At this stage the load increments are necessarily small (2.0 psig) in order to accurately predict the cracking pattern and stress redistribution.

Thus, it becomes possible to detect if a failure mechanism mav develop at the base of the cylinder before yielding of the primary membrane reinforcement.

4 i

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3-5 Based on the results of the analysis described above, at a pressure of 154 psig yielding of the internal layer of hoop reinforcing is initiated. At a pressure of 157 psig the yielding reaches the external layer of hoop reinforcing and extends over a large portion of the cylinder wall. Up to this pressure the analysis does not predict a shear failure at the base. At pressures above 154 psig it is observed that as the rate of growth of the radial displacements with internal pressure increases, the shear force also increases at a higher rate. Thus it appears that a shear failure at the base may develop at a pressure slightly above 157 psig. Since general yielding has been reatned the load increments necessary to continue predicting containment response must be decreased even further than described above (less than 0.5 psig). Consequently, it was decided not to continue the analysis any further.

l l

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3-6 3.2 Behavior at Large Deformation As discussed in Section 3.1, the containment structure is predicted to reach a general yield state at a pressure of 157 psig, which confirms the estimate provided by SMA. As the pressure is increased above this level the containment structure will begin to undergo large deformations. SMA evaluated the behavior of the containment structure at such pressure levels" and the results of this evaluation are summarized in Appendix H.1 of the PSA2, The hand calculations performed by SMA and used for the probabilistic assessment primarily identify several possible weak places in the structure and determine the corresponding maximum pressure capacity in search of the controlling f ailure mode. The uncertainties in the results are estimated and identified as coefficients of variation (CoV) to account for both uncertainty and randomess of material behavior and lack of knowledge regarding the exact structural behavior. The break of the liner plate is defined as the failure mode. Tne capacity of the containment structure is computed in terms of the internal accident pressure it can withstand. Any leakage associated with the pressure level is estimated with a CoV.

Accident scenarios are postulated for both wet and dry containment condi-tions. The corresponding containment liner temperatures are 271*F for the wet case and 700.*F for the dry case. The structural calculations are first per-formed for the wet case and then modified to reflect the reduced material strengths for the dry case. The various failure modes considered in this analysis are discussed and evaluated below. During the course of its review, BNL observed that the SMA calculations did not show any checker's signature.

As a result, PSNH has committed to perform a complete and independent check for all containment strength calculations.

3-7 Membrane Failure The cylindrical wall and the hemispherical dome are assumed axisymnetric and they take the pressure load by membrane action. Both the hoop and the meridional pressure capacities are determined based upon the ultimate strength of reinforcing steel bars, failure cf which will lead to a gross containment failure. Tne median pressure capacities calculated by SMA at 271*F are as follows:

MODE PRESSURE CoV cylinder, hoop tension 216 psig-(governs) .12 dome, hoop or meridional 223 psig .12 cylinder, meridional tension 281 psig .12 The govering hoop failure at 700*F corresponds to a median pressure of 198 psig. Tne above capacities are based on the assumption that the membrane forces are resisted by the reinforcing bars and the liner plate, and not by concrete.

l Since the above pressure values correspond to the ultimate strength of reinforcing steel (109 ksi at 7.5% strain), the containment will undergo a great amount of expansion before failure. This is illustrated in Figure 3.4 which plots contai7 ment pressure vs. radial displacement of the containment wall as calculated by SMA. At 216 psig the radial. displacement of the con-tainment wall away from the base is in excess of 3.0 feet. SMA believes that at this pressure there is a 50 percent chtnce that the containment liner will remain intact and there will be no gross containment rupture. BNL believes that at these large containment deformations it is difficult to accurately predict the behavior of the containment and that containment liner failure is much more likely. It is also noted from Figure 3.4 that at a pressure of 216

3-8 psig the pressure-displacement curve is almost horizontal. Thus, any further pressure capability of the containment would have to be attributed to even greater material strength of the reinforcing steel. Although some reinforcing steel may have a greater strength, BNL believes that for the high strain levels being considered that further consideration pust be given to instances of progressive failure of the reinforcing steel.

In the light of the above discussion, BNL considers the 216 psig pressure capacity predicted by SMA to be an upper bound failure pressure. BNL believes that a more suitable median failure pressure should correspond to the pressure level at which the reinforcing steel reaches 1 percent strain (175 psig for the Seabrook containment). Such a level recognizes the ability of the con-tainment to withstand pressures beyond the general yield, but limits the amount of containment deformation to levels more commensurate with the current state of knowledge concerning containment performance. It is of interest to note that if Gne assumes that the 216 psig level (ultimate strength).repre-sents a 95 percent probability of failure and the 150 psig level (yield strength) represents the 5 percent probability of failure, then the median f ailure pressure assuming a lognormal distribution is 180 psig. This pressure is more consistent with the pressure corresponding to the 1 percent strain i level.

The median failure pressure corresponding to the 1 percent strain level for the dry condition is 158 psig as indicated by PSNH in the response to NRC question 20 (PSNH letter dated October 31,1986).

Shear Failure of Well at Base The shear failure of the cylindrical wall is estimated by SMA at a median pressure value of 319 psig with a CoV of 0.29. This pressure value is deter-mined based upon the yield strength of the reinforcing steel and on the

3-9 assumption that the critical section will occur at a distance of 0.7 x effective wall thickness above the base. The shear failure corresponding to the ultimate strength of the reinforcing steel is estimated by SMA at a median pressure of 408 psig with a CoV of 0.3.

SMA assigned a large variability to this failure mode due to their uncer-tainty about the applicability of their elastic analysis when some yielding occurs. However, BNL feels that this failure mode is more critical than assumed by SMA. As discussed in Section 3.1, BNL investigated this mode of failure by means of a non-linear finite element analysis. It was confirmed that such a failure is not expected to occur for pressures up to 157 psig.

However, BNL believes that a shear failure mode at the base may develop at a pressure slightly above 157 psig.

Flexural and Shear Failure of Base Slab The flexural capacity of the base slab is determined by SMA based upon the yield line theory. The median basic capacity is estimated to be 168 psig. However, when the friction and mechanical locking between the base slab and the ring girder of the enclosure building are Considered, the median overall capacity is estimated by SMA as 400 psig with a CoV of 0.?5. Conse-quently, it is concluded that flexural failure of the base slab is not a controlling failure mode. The shear strength of the base slab is also calcu-lated considering restraint from the ring girder of the enclosure building.

The median pressure capacity is estimated by SMA as 323 psig with a CoV of 0.23.

BNL reviewed the SMA calculations concerning the shear and flexural fail-ure modes of the base slab and agreed that these failure modes would not be controlling.

3-10 Containment Daformations The deformation of the containment is calculated by SMA based upon the assumption that concrete will share tensile load with steel even at the ultimate strength of steel. This assumption presupposes a bond between concrete and the reinforcing bars up to the failure pressure except at the location of the cracks which are postulated to occur with a spacing of approximately 21 inches. The biaxial tension test results presented by Julien and Schultzs indicate that concrete will crack at an early steel stress level and the deformation at a high steel stress level is due to steel strain only.

The hoop and the meridional reinforceing bars used in these tests were of same diameter as those for Seabrook, namely, No.18 and No.14 The concrete cracked at a steel stress of 9.4 ksi and the effect of concrete stiffness dis-appeared beyond a steel stress of approximately 25 ksi. Consequently, BNL is concerned that SMA may have underestimated the containment deformations corre-sponding to the containment pressure levels. Since .the containment deforma-tions can result in containment penetration failures, an underestimation of the deformations would result in higher oredicted failure pressures for the penetrations.

In response to this concern PSNH provided a comparison of the containment pressure displacement curve with and without bond stress (RAI 32, PSNH letter dated November 7, 1986). This comparison is shown in Fig. 3.4 PSNH con-cluded that an assumption of no bond stress would have no effect on the con-clusions of their studies. The effect that this assumption has on the reported capability pressures for critical containment penetrations is dis-cussed in Section 3.3.

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3-11 3.3 Capability of Penetrations Many penetrations through the containment shell are provided. These include a few major penetrations such as the equipment hatch, personnel airlock and fuel transfer tube and numerous smaller penetrations accommodating system high energy piping, moderate energy piping, electrical, instrumentation and ventilation lines. All penetrations are anchored to sleeves which are-embedded in the concrete containment wall. For the major penetrations, the containment wall is thickened into a hub around the penetration sleeve with the wall hoop and meridional reinforcing members directed past the opening in a continuous fashion and additional reinforcement provided as sleeve anchor-age. For each high energy line, the penetration is a forged member, termed a flued head, which forms an integral part of the piping and the containment sleeve which is welded to the containment liner. For all other penetrations the closure is a flat plate welded to the containment sleeve and either welded or connected with a compression fitting to the penetrating element. These flat plateclosures accommodate either single or nultiple penetrations.

To assess the capacity of large penetrations, SMA performed an evaluation of the equipment batch. This hatch is the largest of the large penetrations and was considered to represent the bounding or most critical penetration in this category. In the evaluation, the capacity of the hatch anchorage was determined to be in excess of 300 psi. Possible failure of the liner at the hatch juncture due to sleeve-concrete separation was also evaluated and found to be improbable due to the low magnitude of the predicted liner strain.

These evaluations are considered acceptable.

Although the capacity of the fuel transfer tube anchorage was established in the equioment hatch evaluation, the containxent Wall in the vicinity of this penetration is subject to punching shear failure since it makes hard 1

1

3-12 contact with the fuel transfer building when the containment expands. Using a simple approximation to model the loading and relying primarily on doweling action of the containment reinforcement to resist the load, SMA determined a mean capacity of 320 psig in this failure mode. Acknowledging the approximate nature of this calculation, SMA assigned a large factor of uncertainty to the results. Probabilistic aspects notwithstanding the crude nature of this calculation warranted further verification of the results. Therefore, BNL performed additional calculations for this failure mode to form an independent assessment of the important force-displacement parameters.

In the BNL evaluations an approximate model of the system was again used but this model differed from that used by SMA. The results, although differ-ent from those developed by SMA, indicated that no gross deficiencies existed in the SMA calculations. Further, the estimate of the pressure at which contact is made by the containment shell against the fuel trans'fer building, a controlling parameter in the evaluations, is not subject to the large uncer-tainty associated with the force-displacement parameters mentioned above.

Consequentially the SMA calculaticns although approximate in nature are con-sidered sufficient to characterize the impact this failure mode has on con-tainment integrity.

To assess the capacity of small pipe penetrations, SMA performed evalua-tions for three specific penetrations, X-26, X-28 and X-23. X-26 was stated to be a bounding or most critical example of a single pipe moderate energy penetration, X-8 a bounding case for a high energy penetration and X-23 a bounding case for a multiple pipe moderate energy penetration. For each case, simplistic inelastic analysis methods were used to estimate the forces developed at the pipe / penetration interface as a function of containment i nte rnal . pressu r,e.. This data coupled with estimates of the penetration f ailure characteristics allowed the calculation of the probability of penetra-tion failure as a function of containment pressure in each case. The median l

3-13 f ailure pressure and the associated median leak areas were 166 psig/0.51n2, 2

180 psig/50in and >216 psig/61n2 for X-26, X-8 and X-23 respectively.

The discussions included in the SMA evaluations provided the basis for the SMA contentions that the penetrations evaluated were the bounding cases for the penetration types considered. Those discussions, however, did not adequately characterize all other penetrations. For this reason SNA was requested to compile a list of all penetrations, categorize them in accordance with design features and demonstrate that the performance of each is adequate-ly represented and bounded by the sample of three evaluated. As a response SMA provided Table 3.6 characterizing all penetrations and the calculations considered to be pertinent for their qualification.

A review was made of the evaluations provided for the bounding cases. In each instance the structural aspects of the calculation seemed appropriate, with the exception noted below, but the assignment of leakage area was consid-ered arbitrary. In addition, for each case, failure was induced by the displacement of the containment shell. Since the correspondence between this displacement and tne containment pressure is dependent on the bonding assump-tion made for the containment reinforcement and since BNL has requested SMA to perform evaluations corresponding to a no bonding assumption (see discussion in Section 3.2) BNL elected to further assess the failure pressure and leakage area for the two penetrations X-8 and Y.-26 Penetration X-23 was not con-sidered since it exhibited a high failure pressure.

For the hig% energy penetration, X-8, SMA estimated the median failure pressure to be 180 psig for the wet case with an associated median leakage area of 50 in2 and a legnormal standard deviation of 0.5.

The estimate of the median leakage area was based on an annular gap of 1/2 in, for the full circumference, at the containment sleeve. The estimate of the standard devia-tion was arbitrary. For the 'no bond case BNL estimates the median failure

l 3-14 pressure to be 167 psig for the wet case and 152 psig for the dry case. BNL accepts the SMA estimate for the median leakage area but disagrees with the assumption regarding the standard deviation. In the absence of more explicit data concerning the behavior of penetration sleeves at failure, BNL believes that an upper bound for the leakage area approaching the total annulus between the pipe and containment sleeve should be considered. Based on these considerations failure of this penetration corresponds to a type B failure for the median leak and a type C failure for the upper bound leak (these agree with the SMA findings).

For the moderate energy penetration, X-26, SMA estimated the median fail-ure pressure to be 166 psig for the wet case with an associated median leakage area of 0.5 in2 and a lognormal standard deviation of 0.69. The estimate of  !

the median leakage area was based on an annular crack of 0.06 in, the machined clearance between the pipe and the thru hole in the closure plate, extending over 607. of the circumference. The standard deviation was derived by consid-ering 0.02 inches a miniumum crack width and full circumference cracks. For the no bond case it is estimated that the median failure pressure is 159 psig for the wet case and 147 psig for the dry case. Regarding leakage area, the estimate for the median leakage area is accepted but an assumption for the upper bound leakage area equivalent to that recommended for X-8 should be used. Specifically, consideration should be given to an upper bound for the leakage area approaching the total annulus between the pipe and the contain-ment sleeve. Based on these considerations failure of this penetration corre-sponds to a type A failure for median leak and a type B failure for the upper bound leak (the SMA estimates yielded type A for both conditions).

As noted above, one deficiency was noted in the structural evaluations for the penetrations. In those evaluations only a simple concrete shear cone calculation for a generic case was provided to show that the p'e netration anchorage capacities were adequate. Owing to the highly cracked state of the l

3-15 containment wall at high containment overpressures the reliance on normal concrete action was questioned. SMA was requested to provide additional calculations to demonstrate that small diameter penetration sleeves do not punch through the containment wall under the worst pressure conditions assumed in the analysis. The applicant's response to this request is currently under review by BNL.

Another potential failure mode for the piping penetrations, is the fail-ure of the pipe both inside and outside the containment. This failure mode was evaluated by SHA for the piping in the sample considered most prone to this failure, the piping passing through penetration X-8. The calculations indicated that the piping failure pressure exceeded the penetration failure pressure. Further given the high ductility of the piping, any failures of the piping would have gross distortion, crushing and section collapse associated with them limiting the size of the potential leakage areas. These evaluations seem appropriate for the piping considered.

Other piping penetrations involve the containment ventilation and air purge systems and the containment sump system. The containment ventilation lines have isolation valves both inside and outside of the containment. For these penetrations, the most likely mode of failure is considered to be deterioration of the valve sealant materials at elevated temperatures. In the event of the seal failure of the inner containment valve, the volume between the valves must fill and achieve an elevated temperature before failure at the outer isolation valve can occur. The elapse time for this failure mode is anticipated to be long as compared to other containment failure modes and is therefore considered of little consequence.

The sump system penetration is at elevation -31'6". A review of the drawings originally provided to BNL indicated that the penetration sleeve is welded to the liner at the inside of the containment and to the train AaB sump

3-16 suction valve containment tank on the outside of the containment. As such the sump suction valve containment tank was considered to be a direct extension of the containment vessel and would have to have sufficient capacity to withstand the temperatures and pressures associated with containnent overpressurization events. It appeared that SMA did not consider the tank in their evaluations and therefore, they were requested to perform an assessment of the capacity of the sump containment vessel to the accident conditions. In the response PSNH provided drawings that showed that the tank is isolated from the containment atmosphere by a welded plate closure between the penetration sleeve and the suction line piping. Because of its isolation the sump tank is not subject to accident conditions of pressure and temperature and no further evaluation of its capacity is required.

Another type of penetration is the electrical penetration' assemblies.

The applicant has indicated that they briefly reviewed these penetrations and that they were not a controlling mode of failure. These types of penetrations have been included in the ongoing SANDIA test program sponsored by the NRC.

B N.'

is currently reviewing the available test data from this program to assess the extent to which the test results will support the applicant's conclusion.

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1 3-17 I l

3.4 Summary of Structural Findings Based on its nonlinear finite element analysis of the Seabrook contain-ment BNL concludes that shear failure at the base of the cylindrical wall is a critical failure mode but would not occur before reaching a pressure of 157 psig.

BNL agrees that the containment structure would reach a general yield state in the hoop reinforcing steel at a pressure of 157 psig and that it is appropriate to consider this pressure as a lower bound pressure for the hoop mode of failure. However, BNL believes that the median hoop failure pressure should correspond to th9 one percent strain level in the reinforcing steel, which is a pressure of 175 psig. The above pressures are for the wet containment conditions. For the dry containment conditions the corresponding median failure pressure is 158 psig and the lower bound pressure (general yield) is estimated to be 145 psig. This latter value is based on the reduction factor recommendation in Section 11.3.4.1 of PLG-0300.

With regard to containment penetrations, BNL believes that the failure pressures should be based on containment deformations assuming no bond strength between the reinforcing steel and concrete. Based on this assumption BNL estimates median failure pressures for the wet containment condition of 159 psig and 167 psig for penetrations X-26 and X-8. For penetration "-26 BNL agrees that a Type A -(less than 6 square inches) leak path is appropriate for the median estimate; however a Type B (6 square inches to about 0.5 square foot) leak path should be considered as an upper bound estimate. For penetration X-8, BNL agrees that a Type B leak path is appropriate for the  !

medium estimate; however, a Type C (greater than 0.5 square- foot) should be considered as an upper bound estimate.

, a 3-18 For the dry containment conditions, BNL estimated the median failure pressures for penetrations X-26 and X-8 to be 147 psig and 152 psig, respec-tively. These values are also based on the reduction factor recommended in Section 11.3.4.1 of PLG-0300.

Although BNL has performed some independent calculations to support its conclusions regarding the containnent strength, it also relied on the results of calculations performed by PSNH and its contractors. Therefore, BNL recom-mends that a complete and independent check of all relevant containment strength calculations be performed by PSNH. PSNH committed to such a check in their letter to the NRC dated October 31, 1986.

3.5 References

1. Containnent Design Report For Public Service Company of New Hampshire.

Seabrook Station Unit Nos.1 & 2, by United Engineering Constructors Inc. , Janua ry ,1985.

2. Seabrook Station Probabilistic Risk Assessment, Pickard, Lowe and Garrick, Inc., PLG-0300, Appendix H.1, December, 1983.

3. Sharna, S., Wang, Y. K. and Reich, M., " Ultimate Pressure Capacity of Reinforced and Prestressed Concrete Containments", NUREG/CR-4149, May 1985.

4 Hand Calculations by Structural Mechanics Associates (SMA), Originated by RP, dated December 1982.

5. Julien, J.T. and Schultz, D.M., Tension Test of Concrete Containment Wall Elements, Transaction of the 7th International Conference on Structural Mechanics in Reactor Technology, Vol. J, pp. 237-244

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i i 1 1

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                                                              ,----.ev,        ,   ,m    a ,   sw.            - - - - - - - - .    - - - - - -   -r--   -- -,<v-

I 3-2E e Taria 2." F ein'ecce eat Details c' tne Containmect De-: i i Hoop Meridional Seismic I Elevation (Both Faces) (Both Faces) Dianoaal 0*(S.L.) 9.4* 2-818 0 12" #18 0 12" #18 0 22" 9.4* 30' l 410 0 12" #18 0 12" #18 0 22" l l #12 0 22" l l j (alternate) i 45' 30" *18 R 12" #18 9 10.4" #14 0 19" i a** EC' ' ele 0 12" #18 0 12" --- 60* 79.2* Al8 012" #18 9 12" --- 79.2* 90* Ale 0 6.4" #18 0 6.4" --- i i y i l 1 Ji i I, i i f

   . _ . _      . . _ . _ - , _ . - _ , , . , . _. _ - ~ _ . _ _ . _ . _ _                           _ , , . . - . . , _ . _ _ _ _ . , . . , , _ . , . .          ,  _

J 4 TABLE 3.4. Statistics of Concrete Compressive Strength 20-Day Old Cylintfors Aqed Concrete

• Median Logarithmic - Merlian Loqarithmic Strenrith St inef arrl Stranqth Standard Concrete Type (psi) D,viation (psi) Deviatinn 3000 psi Design Strength Concrete 4750 0.14 5700 0.17 4000 psi Design Strength Concrete y f or Containment 5450 0.10** 6540 0.14  %

4000 psi Design Strength Concrete for Tunnels 5780 0.0')6 6940 0.14 4000 psi Design Strength Concrete for Other Structures 5590; 0.10 6710 0.14 Median strength and logarithmic stanrfard deviation are obtained by multiplying the 28 day strenqth bv.a rindom factor, which is assumrd to be independent of the 20 day strenqth and has a merlian of 1.2 anti a inrinormal standard deviation of 0.10.

   ** This mmber was estimated..

                                                                                                                                                   .A
   . e 3-27                                                                   -

Ta?le 2.! Concrete Properties l MATERIA; PARAMETER. f'c=3000 psi f'c=4000 psi Young's Modulus 4340ksi 4650ksi Poission's Ratio 0.19 0.19 Yield Strength in Uniaxial Tension 0.233ksi 0.262ksi - Yield Strengtn in Uniaxial Compressior. 2.46Ksi 2.81ksi Yield Strenstr. in Eta >ial Compression 2.85ksi 3.27ksi l Fracture St rer.; - + - t - .= > ' Tersier 0.54ksi 0.61xsi Fracture Strength in Uniaxial Conoression 5.7ksi 6.54ksi Fracture Strain in Tersior 0.00045 0.0005 Fracture Strain in Compression 0.005 0.005 l P t l e b 1 f .I, I t i  ! i  : I i f h I k 1

                                         -. .. -.               . _ ~ . .                 - . - .      _ . . . - _ . -  _ -    _ - - . -         _

TABLE 3.6 CHARACTERI7AT!0fl 0F C0flTAINf4ENT PfflFTRATIONS 3 7_--___-.- _ - _ _

,                                                      Penetration Closure                        l'enet ra t ion flamber                       Penetration Specifically i                                                                Tvpc                                                                                                                          Qualification Analyzeil                             tf..t hoit i

I. Flueil 11ead X-1 to X-8, X-0 to X-15, X-8 Iteport pa ge r. f t .1 4 '. to X-63 to X-66 (IR inch, sch 100 carbon Steel) l. 11 . 1 '. 0 II. Flat Pla te Closure X-25, X-26, X-77 X-7 f, 11,i c k tJa l l - I.a r ge Page 11 . I - 3 9 ta II .1 -4 ?. Is..r e Piping (4 inch, sch 160, stainless) III. Flat PlateClosure X-16 thru X-24 X-23 rage s II.1-39 t . ti .1-4 3 Thin Wall - IArge X-28 thru X-34 (I? inch, sch 40, Carbon Stect) Bore an.1 Small Bore X-39, 4I, 42, 50, 60, Piping 61, 67 IV. Flat Plate Closurc X-35 t hru X-38 X-71 Thin Wall Piping X-40, X-4 3, X-4 7, X-48, Page 11.1-37, 11.1 19 Pfaltiple Penetration X-49, X-50, X-57, X-51, X-71 tirro X-76

v. Fuct Transfer Tube X-62 X-62 Page II.1-50 t o 11. I ' 5 m m.

i 4

a-1 4 COU4."E* EVE *. TDEE REVIE/ In tnis sectio- the ability of the Seabrook containment to contain severe accident loads is exanined. Note that ways in which the containment might be bypassed or not isolated are discussed separately in Section 2. This section, therefore, specifically deals with ways in which severe accident loads might result in loss of integrity of the Seabrook containnent. The section is divided into two major parts. Firstly, che impact of uncertainties in con-tainnent locds is examined. Secondly, the impact of the BNL review of the contain:re . beha.ior (in Section 3) is assessed. Finally, the results of the section are sun :arizec in Section 4.3. The botton line of the updated assessment of containment performance for Seabrook is given in Table A.1, which was reproduced from PLG-0465.I The conditioW probability o' a gross early containnent failure given a core reh accident was predicted to be 0.0C at Seabrook compared with 0.34 in the RSS.2 Tne probability conditional on core melt of early failure in PLG-0465 is an order of nagnitude lower than in the SSPSA.3 This is largely due to the redu:ed f reque .0, (relative to the value in the SSPSA) of interf acing systen LOCAs, which are discussed separately in Section 2. Ncte that the containrent event tree quantification in the 55D59 was reviewed at BNL in NUREG/CR-4540. The BNL review was limited in scope and did not include at that time a detailed assessment of the Seabrook containrent behavicr. Tnis has subseque".tly been performed as part of the present revie.., and it is docunented in Section 3. However, the review of the SSPSA was sufficiently detailed to allow BNL to conclude:

           "There is negligible probability of prompt containment failure. Failure                       l I     during the first few hours after core nelt is also unlikely and the timing of overpressure failure is very long compared to the RSS. Most core nelt acci-dents would be effectively mitigated by containn'ent spray operation."

The above conclusions were not based on Seabrook specific calculations perforned at BNL but reflected our best judgment based on extensive reviews of i

4-2 othe* SiPile" co"tair"t"! desig"? (ir particular, our revie.'I c' thi Zic" Probabilistic Safety Study6 ). In this section, we critically review the abcVe conclusions basec on our current understanding of containment loads a"c

,       performance during severe accidents.

4.1 Sensitivit:. tc Centtinrer.: Leads During a core melt accident, there are several possible types of contain-ment loads tnat could occur. Each are briefly discussed below: He combust 1c": Du*ing a core r.elt accident, significant quantities of H: and other combustible gases coulc be generated. If these combustible gases accumulated to large concentrations before igniting, the resulting deflagra-tion could impose a high pressure /tenperature load on the containment. The applicant preser.ted informatior tc incicate that such loads would not serious-ly challenge the Seabrook containment. Stean/noncondensible cas partial pressures: Without the containnent heat renoval syster.s operatir.g, stea and noncondensible gases generated during the core celt accident WoJld cause the pressure in Containment to increase. At the tire of vessel failure, there is the potential for the hot core debris to contact water. This contact co;1d result in rapid boiling of the water and a sharp pressure pulse in containment. Limiting calculations were performed to demonstrate that the pressure pulses resulting from simply boiling the water woulc not pcse a tnrelt tt the Seabrook containment. 1 Direct contact of core debris: In some containment designs the contain-I ment boundary is directly accessible from the region below the reactor l vessel. In these designs the core debris after it melts through the reactor vessel could contact the containment boundary. However, the Seabrook contain-j nent design is not susceptible to this mode of containment failure, j Stear exolosions: When molten core materials fall into water, experi-rents indicate that the boiling can become explosive in nature. It has been postulated that these explosions could generate missiles which could directly f ail the containment boundary. The potential for an invessel steam explosion e

c-3 to 0: cur an- g="c-ate a rissile carnic o' failing a certaia f ** buildia; wH investigated by a group of experts, and the results were published in NUD.E*- 1116.7 Tne cenclusio- o' this expeat group was that such events have a rela-tively low probability. The results of this expert group are consistent witn the applicant's submittals on Seabrook. The allocation of a very low proba-bility (10 " cead4ticaal en cere nelt) to this event is supported by the authors. 4 Dire:t containment heatinc: Tnis is an area of significant phenomenolog-ical uncertainty related specifically to core meltdown with the primary syster 3 at higt pressure. If noiten core materials are ejected from the reacter vessel unter pressure, exper1 rents! at Stil have indicted that they form fir:e aerosols, which could be dispersed into the containnent atmosphere and direct-ly heat it. An additional concern is the oxidation of the metallic content of f the core debris. These reactic .s are very exothernic and would add an addi-tiona? heat load to the cc9tainrer.t. The pressure rise in containnent dus t; direct heating is directly proportional to the quantity of core debris eje:te: from the reactor vessel and to how nuch of this core debris is dispersed into the coa.tainreat atmosphere. Tnt applicant considered that this pheno ,er,a was  ; n t a concern at Seabrook be:ause c# the design configuration of the contain-nent, which they felt would prevent dispersal of the core naterials into the bulk o' the containnent atnesphere. The combined probability (conditional on core melt) of the above phenone-na resulting ir ea*.; contattrent f ailure was determined by the applicant t; be less than 10 " for Seabrook. Within the scope of the present review, BNL l has not performed Seabrook-specific containment loads. However, BNL has been involved in updating' the risk profile for the Zion plant for input to the NRC's " Reactor Risk Reference Document," NUREG-1150.10 1 I qhe updating of risk for Zion was based on a methodology developed II as part of the Severe Accident risk Reduction Progran (SAN 086-0119) at SNL. This methodology used expert judgment in an attempt to estimate the uncertainty associated with some of the phenomena noted above. The methodology also addressed other areas of uncertainty such as accident sequence probabilities, source terms, and containment perfornance. The methodology was developed at i

4* S*u speci'itally *cr tre 5;-ry riant but was extrapolated te Zion at E'J., Tec Zion plant is very sirilar to Seabrook in terms of the containment volune te reactor power retic. inus, extrapolatior of the Zion loads to Seabrook would give some indication of the impact of applying this new methodology to Sea-brook. It must be emphasized that this exercise should in no way be inter-preted as a Seabrook-spec 4'ic calculati0n because the conposite containment failure probability distributions for the two plants appear to be quite dif-ferent. It simply gives some indication of the sensitivity of the Seabrook results to the types of uncertainty discussed in NUREG-1150. It should also 1 be noted that this work 15 p e,;ainary and has not yet undergone full peer review outsice of M: a c its cortractors. it is, therefore, subject to revisio.. However, applicatic" of the new methodology to the Zion plant gives a r.edi an probability c' app c>in tely 10-2 for an early containnent failure conditioca' on core relt. Tre higher probability of early containment f ailure { is dJe to Dore Conservativ6 assrptions about direct Containment heating and ' H2 conbustion taken in combination with high steam /noncondensible gas partial pressures. These assu ptic : u rt considered by the applicant to te r:ct applicable to the Seabrook ccntain ent. However, the applicant was requested { at a reeting on Noverber 12, 19FF to assess the impact of the NUREG-1150 assumptions race for Zica c tre ecse versus distance curves and the risk prc. files in PLG-0465. In reference 12, the applicant provided a response to this j infornation request. The cerclusicas given in reference 12 are quoted below: (1) The NUREG-1150 assumptions with respect to early containment failure are not credible in the context of the Seabrook specific design and configu-ration. (2) Postulating these assunptions in the sense of a sensitivity analysis would not have a large impact on the results presented in the WASH-1400 sensitivity study. (3) All of the conclusions with regard to emergency planning would still be { valid. Tne calculations that forn the basis for the above conclusions were also presented in reference 12. These calculations were reviewed at BN' and (given 5 l

4-5 tne assu--tic's i-'eser' we'e found to be correct. Tne increases in risk estinates with anc without evacuation were found to increase by relative srall amounts and we*e still less than the p r;:cd NRC safety goals for individual risk and societal risk. The inpact on the dose versus distance curves was not presented in reference 12 but it was assumed to be small. The results in reference 12 are based c- ?" applicar.t's assumption (refer to Figure 4.1) that for a conditional probability of early failure at Seabrook of 0.01 most of the failures will be type B leaks rather than gross type C failures. This implies that nost cf tne early f ailures will result in an $6W release rather than the nore severs S;W release. The appropriateness of this assunption was revieaed by E'C. sta" in Se::ien 3. Tne impact on risk of the BNL review is given in tre next suosecticr.. 4.2 Sensitivity to Containreat Perforrance The cc pesite ccr.tainre .t f ailure probability distributions for various types of cor.tainment faiiv e are given in Figure 4.1, which was reproduced fron reference 12. Two total f ailure probability distributions are given for

                           " dry" and " wet" cc-d'tions. T*c "wct" failure distribution corresponds to

accidert sequeates ia wnich the reacter cavity is flooded, the core debris forms a coolable debais bed, and all of the decay heat goes to boiling water.

This configratica is a relatively effective way of pressurizing containreet (i f the containment heat removal systens are not working). However, the temperatures in the containrent are relatively benign and would be Close to sa t u ra t i cr.. Tre ' cry' 'tilure distribution corresponds tc accident sequences in which the reactor cavity is dry and extensive interactions occur between the core debris and concrete. Under these circumstances, hot noncon-densible gases are generated during the core / concrete interactions. These gases result in slower containment pressurization than the " wet" condition but the hot gases heat the containnent atnesphere to higher tenperatures. Thus, for any given pressure level, the "dev" accidents would have higher atmospher-ic temperatures than the " wet" accidents, which in the applicant's analysis (refer to Figure 4.1) resulted in a higher probability of containment failure, 1 in orcer to assess the potential for early containnent failure due to I short duration pressure pulses, we considered it appropriate to use the " wet" l l l l l

   ,-_-,___.__,_m.__-._                        , , _ _ _ _ , _ _ . , _ , , . ___ , _ . . _ _ . . _ . _ . , _ _ _ _ . . . _ . _ _ _ _ , . . - ,   __z., - _ _ . -     . _ _ . . _ _ _ . -

4 l 4-E pr:baM !ity C it ** r d i:' , TM s is sir;.ly because the co". tai- e-t structuet: would not have bee exp0 sed to the higher " dry" atmosphere temperatures prio-tc the pressure f alse occurrirg. Tne structures would therefore not have degraded as a result of high temperatures and the " wet" conditions are more appropriate. If we, therefore, focus on the " wet" failure distribution (in Figure 4.1) t as being more appropriate to determine the potential for early containment failure, tnen tne relative contrioutions of gross (type C) failures and benign (type B) f ailures at a conditional probability of 0.01 were determined 12 by the applicant (refer to Ficure 4.1). The impact of the above assumptions or the dose versus distance and risk profiles were presented in Section 4.1. I r. this section we assess how the BNL review of containment behavior (refer to

Section 3) night inpact the above results.

l The recian f ailure pressure based on koop failure assuming " wet" conoi-tions was estirated to be 175 psig in Section 3. In addition, the possibility of shear f ailure at the base of the containment at a pressure of 140 psig was also discussed ir. Section 3. How.ever, tne median probability of approximately 10 2 for a large early containment failure at Zion (refer to Section 4.1) was o. based on W failure pressure of 134 psig. It is not therefore likely that the Seabreak spe:ific failu e pressu es (estirated in Sectica 3) would significantly change the conditional probability of early failure at Sea-brook. In fact, as the Seabrook failure distributions are higher than the Zion failure distribution, one woule expect tne conditional probability of early failure to decrease if everything else were the same between the two pl a nts . However, the Seabrook specific failure pressures discussed in Section 3 are lower than the distributions presented by the applicant in Figure 4.1 and i closer to the Zion values. This, in turn, affec,ts the assumption nade by the applicant, in reference 12, related to the relative contributions of gross (type C) failures and benign (type B) f ailures at a conditional probability of 0.01. Tnerefore, because of the uncertainty associated with the relative con-tributiors of the two failure modes, a limiting sensitivity calculation was perforned to assess the impact of assuming that at a conditional probability 3

4-7a i l of 0.01 all failures were gross (type C) failures. The results of the sensi. ! tivity arc presented in Table 4.2 and Figure 4.2 l 4.3 Sume ry As a sensitivity study, the impact on risk of more conservative assump-tions with regard to containment loads and perfornance during severe accident conditions has been made. The results of the study are shown in Figure 4.2 and Table 4.2. The sensitivity study resulted in a higher conditional proba-bility of early containment failure than considered credible in PLG-0465 (10-8 l versus 1C**). The higher probability was not based on Seabrook specific calculations but inferred from calculations performed at BNL for a nuclear plant (Zion) with a similar containment volume to reactor power ratio. It must therefore be emphasized that this sensitivity study is not $eabrock-specific and it was performed simply to assess the robustness of the PLG-0465 results to uncertainty in containeent loads and performance. The impact on risk of considering a conditical probability of 10-8 was found to be not significant when the composite containment failure probability distributions developed in PLG-0465 (and reproduced in Figure 4.1) were used. Tnis was because, at a probability of 10 8, most of the containant failures would be leakage f ailures rather than gross failures. A leakage failure results in lower offsite health effects than a gross failure. However, if the j nore conservative BN'- assumptions regarding containment perforrance (refer to

                   '$ection 3) are combined with the higher probability of early containment failure then the dose vs distance and the risk profiles in PLG 0465 were affected. As a bounding sensitivity study, it was assumed that all failures at a conditional probability of 10 a would be gross failures and the results i

are given in Table 4.2 and Figure 4.2. The ispect on the dose vs distance curves is similar to the influence of the more pessimistic assumptions regard-i ing induced steam generator tube rupture (refer to Section 2.3 and Figure 2.4). The similarity of the results is to be expected as, given the limita-tions of the RSS source term methodology, both types of events had to be conservatively binned into the same source term category (namely $1W). The impact on the risk of cancers and early fatalities of more conservative assueptions regarding containment loads and performance is given in Table 4.2 i

  ,_--____-.._,m                ~ ,-_-____ - - - -- ._.

4-7e and again follows similar trends to the $GTR study (refer to Table 2.8). The impact on the risk of cancers is sr'all but the early fatality risk increases by less than a facter of 2 with no 1 mediate protective actions taken. 4.4 References

1. PLG-0465.

2. R$$.

3. $$PSA.

4 NUREG/CR-4540.

5. NUREG/CR-3300, Volune 2, " Review and Evaluation of the Zion Probabilistic Safety Study," July 1983.

6. IPS$.

7. NUREG-1116.

                             $NL Direct Heating Experiments.

8.

9. NUREG/CR-4551, BNL/NUREG-52029, Volume ti, Draft, September 1986.

10. NUREG-1150, to be published.

11. $AND86-0119.

12. PSNK Letter (SBN-1237), detec hovember 21, 1986, " Emergency Planning Sensitivity Study," J. DeVincentis to S. M. Long.

e e t_; 1c _

             ~

TCTAL F AILunt

             -              Patssum t:
                                                                      ..'..'                           # , e - -l 3;;                                                                             ---
             ~

• p 7- / BENIGN (TYPEEl

             ~
                                                       .                  /                 /               PAILUmE= m ET
                                                                      /              /
                                                                                        /                   M OUENCLS

• i

                                            .                                      /
                                       /                                                    Cnoss (TYPE C:

[ / F AILun t . WET

             -                      ,./                                     /               SicutNCES l                                       /
                                .                                       /
             ~              *
                              .                                       /
                                                                   /

/

g ic i --

             =
                       -!-                    0. 0//

e  : I b 2 [ 5 - ! I E I 5 - : I 8-  : I L so ,  ; I _. I

I

              ~!
               -                             I
              -                             I
              ~
                                          /                               v d( 2 X l0 '

1C* ll l 1 1 I

                                  /

j f69 M'E g e

                            /          t                       '               '          f              f           '

140 sto lac 200 27C 240 700 Pmttsyng (psian Figure 4.1 Composite containment failure probability distributions for benig, failure, gross failure, and total failure. (F.e rc:xe d fro . Figure 11.3-14 of SE*;:1237)

49 l l l 1 ! ~. **i a a . . . ...; ,

-l

   -     e, -

w=

                                                                                                                                     )
                                                                                                                 ~
                            ~

EE og

        >u                                                                                                        -

8" e5

                           ~

iL W  ; aa , om 01 NUREG-0396 l E , T z< -

                          ~
                                                                                                                 ~

85 w> - . WG - w ut -

u. u -

C2 -

           *              \
       >h E                  ~\               FLG-0:f5                                                             -

Ec \

       .h[v0.c1 -        -

I C 2 - I, . c< . -

       -w               -
                                 ._I _ s s                                                                    .

s=

       ,e l
                                    ,       g                                                                  '
                        ~

l s Impact of early containnen:

       $3C                                       s s

p failure

                        ,          l
                                                     \      s                                                  .

J I l

0. col I - ' - .

                                                              \     *
                                                                         '   ' ' '  l   ,   , , , , , , ,

lo 1c0 0lSTANCE (MILES) Figure 4.2 Comparison of Ei;L sensitivity study related to early containrent failure with PLG-0465 and NUREG-0396 (200 rem plots with no immediate protective actions). l l t

. o 4-10 Ta:,i c J.. Ccma ist . of Core felt Frecuencies and Distributic-of Reieuse Types (reproauced from TaDie 2.2) of PLG-0 5E; Ri o- 03----*-.

                        - - ---'                      WASH-1400 SSPS,.'

Updated PWR Results* e Mean Core Melt Frequency (events 9.9-5** 2.3-4 2.7-4 per reacter-yea-) e Percer* Contriou:1on of Release Ty pe s. . Gr ss, Early Con:ainmer.: 34 Failure 1 0.1 Gradual Containmen: 66 73 60 Ove : essurizati: c-Melt-inrougn Cc ;ain.en: Ints:: 0 26 40

   *Basec on RME:5 (PLG-0:2 1
   **Basec cn WASn-It.00 un:ertainty ranges.

NOTE: Exponential i.e., 9.9-5 =notatie- is incicatec in abbreviated form; 9.: > IC*t .  ;

                                                                                   \

l

4-11

                                                                                                                                                                                          )

Table 4.? Impa:t of P'c Sansitivity Studies oe. PLG-0465 Risk Estimates Absolute Risk per Reactor Year 4 Early Fatalities Total Cancers 1 Sensitivity One Two One Two No Mile Mile No Mile Mile Study Evac. Evac. Evac. Evac. Evac. Evac. Original 2.7(-3',* 3.E(-4) 2.4(-5) 1.5(-2) 1.4(-2) 9.2(-3) PLG-0465 Results Containne".t 4.7i-3) 2. 4 ' - 3 '; 2.0(-3) 1.8(-2) 1.7(-2) 1.2(-2) Loads and Performance

                            *2.7(-3) = 2.7x10 3 1

l l 1 4 l l i i 1 r - -- , e-,----- -- , , ....--,---,-------v---,-ee- ,--,,..~----,,,,,m-wa. , , , . - - ,m- y- - - - - --. - - . --

o . 5-1

5. REVIEW P SOKE TER"5 In this secticr. the fission product source terms developed for PLG-04EE l are reviewec. Tne source terns used in PLG-04651 are reproduced in Table 5.1. The probabilities of each of the source terns are given in Table 5.2.

1 5.1 Fidelity to WASH-1400 Methodology i The fission product release fractions in Table 5.1 were determined by the applicant using RSS2methodology. These source terrs are consistent with the point-estinate source terrs used in tne ' SSDSA. 3 Tne SSDSA source terrs we e reviewed by BNL in NUREG/CR 454T and they were found to be reasonable giver. the limitations of the RSS methodology (principally the CORRALS code). We- reviewir; the PLG-0455 source terns, questions were raised and trans~itted to the applicant. One question related to release catego y 52'.? (refer to Table E.1). Tne fractional release of the noble gases for $24 was quoted as 0.123 whereas the release fractions for Cs and Te were quoted as 0.2 and 0.19 respe:tive!y. It appeared inconsistent to release more aerosols (Cs and Te) that noble gas, and the applicant was requested to either explain the predictions or provide revised source terms. In reference 6, the applicant provided a response to this cuestion. Basically, the noble gas release in Table 4.3 (reproduced in Table 5.1 of this report) of PLG-0465 was a nis-print. The noble gas release fraction for the S2W-3 release phase should have been 0.23 ratwe than 0.023. However, as the SiW-3 release phase cccuas ve ; late (approximately 20 hours) af ter reactor scram, the impact on risk of this larger noble gas release would be very small. Calculations performed at BNL have verified that the impact cf increasing the noble gas release in the S2W-3 phase on both the dose versus distance curves and the risk profiles is negli-gible. 4 5.2 Credit for Scrubbing of Submerged Releases-Another question related to the credit assumed for the interfacing syste-LOCA source tern in which the break location was assumed to be submerged under l water (57W in Table 5.1). In PLG-0465, the source term mitigation resulting l 1

                                                          ,7    _ . . - . -         -
                                                                                       --rr-._           ,  . _ _

5-2 fro a subcocle: 30 foct deep pool was modelec as a decontarination factcr (0 ) of 1000 for all release fractions except the noble gases. In the RSS a deccntamination factor of 100 was used for fission product scrubbing in a sub-cooled pool. Tnus, in order to be consistent with the RSS methodoloov it. appeared that a lower DF should be used. However, if the pool were indeed subcooled, calculations at BNL indicated that using a DF of 100 rather than 1000 would have no impact on the risk calculations pre-sented in PLG-0465. A core important questien was whether or not the pool would be subcooled or saturated. in the RSS, ne credit was given for fission product scrubbing in a saturated pocl. and therefore, the applicant was requested to provice justificatioS fe" the subcovss assumption. In reference 7, the applicant provided a response to this question. Arguments were provided to indicate that the pool would be at least 1000 subcooled and that this degree of sub-cooling together wi th the large pool depth was sufficient to justify a Dr of 1003 rather thea a D: of 100 as used in the RSS. However, the objective of the questice wts prirarily to oetermine if the pool would be subcooled anc based on the response, this appears to be the case. BNL had already concluded that changing the De fro- 1000 to 100 would not change the dose versus , distance nor the risk profiles. Finally, the conclusion (given in reference

7) that even if pool decontamination were completely ignored, that the dose versus distance and the risk profiles in PLG-0465 would not be significantly ef f ected was examined at BN.. Calculations at BNL indicated that if the fre-quency of interfacing systen LOCAs reported in reference 1 was used, then the coa.clusier was cearect. Tnc arpecpriateness of the interf acing systen LO '

frequency was reviewed in Section 2. 5.3 Sunnary l In summary, the fission product source terns used in PLG-0465 appear in general to be consistent with the approaches used in the RSS. The misprint in Table 4.3 of PLG-0465 related to the fraction of noble gas release was found by the applicant (and confirmed by BNL) to have negligible impact on the risk i profiles or the dose versus distance profiles reported in PLG-0465. Thus, the corrected noble gas-release fraction in the $2d release category would, by itself, not change the conclusions in PLG-0465. In addition, the argunent

o . 5-3 prese-te: by the at:11:ent tntt tre watea in the RHE vault is sufficiently subcooled to warrant consideration of significant decontamination appears reasonable. Finally, the statenent in the applicant's response that even if pool decontamination had bee 9 ignored, the risk profiles or dose versus distance profiles would not change significantly was confirmed at BNL. Note, however, that this Con:lusion it based on the frequency esticatJa for inter-facing system LOCAs in PLG-0465. The influence of the BNL review of interfac-ing system LOCA events at Seabrook is given in Section 2. 5.4 References

1. PLG-0465.

2. RSS.

3. SSPSA.

4 N'JRES/CR-45 0.

5. Corral Code.

6. PS' 4 Letter (SB'i-1234), date: havember 17, 1986, " Response to Request for Additional Inferration (RAls)," J. DeVincentis to S. M. Long.

7. PSNM Letter (SP'.-1227), dated N3verber 7,1985, " Response te Request for Additional Infornation (RA:s)," J. DeVincentis tc S. M. Long.

d Tabic 5.1 Release Categories for Seabrook Station Based on MAsil-1400 Source tenn MethodO10ny

                                                                         ,,,,, ,        Release  Release  Warning   Incrgy                                                           Release Tractinns Time    Duration  Time     Release
                                                                               #9 'Y                      (hours)   (MrA/5)                         0.1.                 I-?          05           TE         B4      RU     IA (hours)  (hnurs)                                   XE 51W           2.5       0.5     1.0      11.9                0.9           7-3              .7            .5             3         06      .07    4-3 5?W-1         4.8       2.0     0.5        0                     01     2.1-4              4.3-3           .023      4.7-3       2.8-3   R.4-4  R.4-5 SN- ?         6.R       4.0     2.5        0                    01      5.0-4              1.1-1          .flin         0 19     5.5-3   3.4-1  5.7 1 57W.1        19.8      18.0    15.5        0                     073    1.6-3              ? .1 - t        .1?6          117      .014     .011 I.9-3 10!At         4.8      24.0     0.5        0                   .l?1     7.3 1              7.9-1          .70           19        .072      015 2.5-3 51W           6.0      24.0     2.0        0                 4.7-4      3.3 6              1. 2 '.       1.7-4       1.5-4       1.9-5   1.2-5  2.0-6 St.W-1        1.75      1.0     1.5       0                     15      1.1-3               .10             11          02        .014   4.13   4.1-4 8.W-7       2.75      4.0     2.5        0                     42     2.9-3               .0/             19          063       .072    .009   .tpil ui 5#W-3        15.75     18.5    15.5        0                   .37      2.2-3                 01          .13          .32        .011    .070  3.n.1   L TOIAL         1.75     23.5     1.5        0                   .9       6.2-3                 18            43          40         047    .033  5.2-3 57W            8.5       7.0     2.0       0                    .9          7-6                7-4           5-4         3-4        65       2-5   4-6 N0fr: Imponentf al notation 15 Indicated in abbreviated form; f.e., 7 3 = 7 x 10-3 a

1 4

                                                                                                                        -+e . . . ,

5-5 ac. i . . . ; - . - n . C *:i . - . > f or *.e<. 5:. ce Te rr Ca te 7:-i .. Plan: Da. age Source Term Category S*3*e e ( f requency) -- se- 3* 5'* 56 57 1F (2.0-5), 1.0

                                                   ,                                                      ( 2.0-8 )

, 1FV 1.0* -- - - - - - - - - - - - - - - - - - - -- - (4.6-9) (4.6-9) 1FP 1." ( 1.4-6 ) (1.;-f' s 1 F. 1.0 ( 2.7-5 ) (2.7-8) 2A 3.4-5 1.4 ; 1.0-2 0.99 (1.9-6) (6.5 *' .. ( 1. 7 -1 ' ' (1.9-6) (1.v-6) > 30/70 2.0-f 6.0-E U.95 0.C: (3.2-5) ( 7 . 6 - l '. ' (3..-?~ (3.6-:) (1.9-6) 3F/7F 1.0 ( 3.0-7 ) ( 3.0-7 ) 3FP/7?? 1.J ( 1. 9- 5,) (1. -:) 4A/SA 3.1-: 1.'-- 5.2-3 U.995 (1.1 0) ( 3 .1 '. ' . ...- g ( .:-7) (1.1--) 7FPV 1,c (1.2-2) ';- 80 1.1-6 ~ 3.1-5 0.9999 (1.0 4) (1.1-10) (3.2-9) (1.0-4) i Total 5. 2 'i 2.0-5 1.4-4 1.1-4 3.2-7 Frequency 3.9-8 N07ES:

1. Exconential notation is incicated in abbreviated form; i.e., 2.0-d = 2.0 x 10-8

2. tium:e s insice pare-: eses are unconci:ional f recuea.:ies (evea.:s per reactor  :

y*ar' tase: On sar valust. fiu :ers nc: ins 1ce : area:neses are con:1:1one!- f re:;en les er s:; :e e- ca:eger,es, given tne in:1cate: l ar.: cacage sta:e. { a ,i s : Sase: cr. ea. values. Me:1a values cf sour:e :ern :a:eg:.-tes are prese .:e: in Se::1:n 3. I I I

                                 -r-T                                                           _   ___.

w ,-

E-!

6. S!Ti CONSEOJEY.E MJ L 6.1 N'SEG-0396 Easis NUREG-0396 3 int roduced the concept of generic Emergency Planning Zones (EPZ) as a basis for the planning of response actions which would result in dose savings in the immediate vicinity of nuclear facilities in the event of a serious power reactor accident. The actions would be triggered if projected radiation doses to an individual would exceed Protective Action Guides (PAGs),

as discussed and referenced in NUP.EG-0396, although ad hoc actions could be taken at any tire. The PAGs are 1 to 5 ren whole body dose and 5 to 25 re-thyroic dcse but are nct intentec to represent acceptable dose levels. Fur. thermore, protective actions may not assure that PAG levels can be prevented. it was concluded in NUDEG-0395 that the objective of energency response plans should be to provide dose savings for a spectrum of accidents that could prodJte offsite dosei ir excess of the PAGs since no specific accident could be identified as the one for which to plan. The nost important guidance for emergency planners is the size of the EPZ. Based on f actors that included risk, probability, and accident conse-Quences, it was juaged that a generic distance of about 10 miles was appropri-ate for core relt accicents. Tne less severe accidents would not have conse-quences in excess of the PAGs beyond this distance, whereas the more severe accidents wculd nnt in ge e-al cause early injuries or deaths beyond this dis-tance. Hence, protective actions were judged to be most useful within this distance. 1 NUREG-0396 used the accident release categories of the Reactor Safety i Study 2 (RSS) to compute the risk of exceeding various dose levels in the absence of protective actions for a spectrum of accidents. The RSS accidents and their median probabilities are given in Table'6.1. Using the original RSS consequence model (CRAC)3 and the accidents PWR1 through PWR7, a 200 rem whole body risk curve was constructed, shown as the heavy line marked 0396 in Figure 6.1. This is the level at which serious injuries and sone deaths can occur.

6-2 However, CE*C cir n?t conoute tne 200 ren risk directly. The authors c' NUREG-0395 had to interpolate to obtain the 200 ren curves. Interpolation was performet for each component accident to obtain the conditional risk given the accident. Then the conditional risk at various distances was multiplied by the probability shown in Table 6.1 and divided by the core melt probability of 5x10 5 per reactor year. The results for each were summed to give the overall risk of the accident spectrum. Each risk was computed using about 100 weather samples from a typical series of New York City hourly annual weather data with the assumption that people would follow normal activities for one day follow-ing arrival cf the first plume to reach their location. That is, people would stay at their original location and receive groundshine doses for 24 hours. BNL recomputed the 200 rem dose vs distance curve using an updated ver-sion of CRAC, called CRAC2, and a more detailed grid. The conditional corpo-nert risks were co'puted directly (no interpolation required) and the results for each o*' the WASH-1493 release categories are shown in Figure 6.1. PEC and PL'97 did not exceed 203 rer. The overall 200 re risk curve was then computed using a core nelt proba-bility of 6x10 5 per reactor year, which is the sun of the probabilities fo-PWR1 - PWR7 given in Table 6.1. The curve gives higher risk for 1-3 niles and lower ris'. for 4-10 niles then NUREG-0306, but the curve still drops sharply beyond about 10 mi l es . It should be noted that most of the core nelt proba-bility comes from pWR7 which does not contribute to the 200 rem curve. It can be concluded that the approxir.aticas used in NUREG-0396 are not substar.tiall 3 different fron the more detailed calculations done by BNL using CRAC2. 6.2 Consecuence Modeling The applicant av used the CRACITs code for their consequence assessnents in PLG-0465.10 In this section CRACIT predictions care compared against conse-quences'nodels currently being used by the NRC and their contractors, namely CRAC, CRAC2, and MACCS.6 The factors involved in consequence modeling are discussed in Appendix 6 of the RSS. All codes compute early and delayed l health effects fron cloudshine, inhalation, and groundshine. Tne early health effects are basec on data f rom the Marshall Islands, bomb tests, clinical data l l l

6-3 f ro: ra:i at : cr. theraa , and lac animals (particularly for lung data). Tre three CRAC models (CRACIT, CRAC2, and CRAC) use a stepwise linear function with a threshold dose for early effects as discussed in the RSS. The MACCS coce (re:ently developed C Sandia National Lab) uses a hazard function approach without a threshold as discussed in NUREG/CR-4214. The latent effects in the CRAC nodels are calculated from the BEIR-1 report e which uses Japanese data plus a modification to the linear dose response curve to account for reduced effectiveness at low doses. MACCS uses the BEIR-3 model 8 which is a linear quadratic dose response model with absolute risk of cancer for sore organs such as bo9e narrow and relative risk for other organs, depending c9 population nakeu;. In addition, CRAC and CRAC2 allow only a one " puff' release of racioactivity wherets CDACIT and MACCS allow "multipuff" releases. There are also other differences in the codes, such as the shape of the plume, dry and wet deposition of particulates, weather sampling, resuspension, etc. which can account for differences of a factor of two in the results. BSL used tne MACC5 coce and the source terms defined in Table 5.1 tc review the calculations presented by Seabrook in PLG-046510 using CRACIT. The comparisons are based on the 200 rem dose probability vs distance curves using the source terr:s and weather data supplied by Seabrook. In addition, BNL cal-culated the individual risk of exceeding 5, 25, and 300 rem to the thyroid and the individual risk of deatt as a function of distance. In all cases, it was assumec tnat the population was exposed to one day of groundshine following arrival of the first plume segment. They would also be exposed to other plumes that arrived at their location within the 24 hours. 6.2.1 Whole Body Dose Vs Distance The MACCS code does not have an organ defined as "whole body" so red marrow was used as a substitute. In CRAC2 calculations it was found that the red marrow dose was about 30*. higher than the whole body dose. Also, early health effects are sensitive to the red narrow dose. Thus, the red narrow dose is a good substitute. . MACCS does not directly calculate the risk of a dose vs distance, and it was necessary to define an oppropriate risk function to obtain this curve. In

 . o 64 the M'::S calculasic 1. tre ri s t o' exceeding 200 re- to the red re-rev: was set to zero if a weather sample yields a mean dose less than 200 ren and one if a weather sa71e yielos a mean dose greater than 200 rer. Thus the mean risk is dependent upon the annual weather data. The risks were calculated for the Seabrook release categories S1w (one puff), S2W (3 puffs), and 56W (3 puffs). The results ere given in rigures 6.2, 6.3, and 6.4.       The conditional probability of .001 risks extend to 25 miles for SIW, to 2 miles for S2W, and to 4 miles for S6W. These distances are somewhat less than those calculated by Seabrook.

6.2.2 Thyroid Dose Vs Distan c - The thyroid doses were not presented by Seabrook in the reviewed report but were discussed at some length in NUREG-0396. .Hence, BNL calculated the risk of exceeding the dose levels of 5, 25, and 30". ren to the thyroid, as was dore in N" EG-0396 f or Seabroot release categories S1W, S2W. and 56W. The results are given in Figures 6.2, f. 3, and 6.4 The results were truncated at 30 miles. The risk of exceeding 5 ren remained above 90% for all three release categories. Tne 30? re curve shows a sharp drop at less than 10 niles for S2J anc at about IE riles for 56W. The curves were obtained by thE same hazard function definitica technicue as discussed in Section 6.2.1. J 6.2.3 Risk of Early Fatalities i MA CS uses ta.e ha:ard fun:ticr. approach to calculate early f atalities as discussed in NUD.EG/CP. 4214 First, the cumulative hazard is calculated as: H = In(2) (D/D5o)* (6.1) where D is the dose and 0 5g is the dose required for producing an effect in 50% of the exposed individuals, and v determines the steepness of the dose 1 effect curve. The fatality risk is then given as: ' Pisk = 1 - e-(HI'H2+H3 + H.,) (6.2)

6-5 whe'e h; is f o* ret re- o. . W: is for lungs, and H3 and Hgare fee the louce large intestine and snall intestine. The risk is assigned a threshold of

    .005.

In CRAC2 and CRACIT. the dose response is piece-wise linear due. to irrad-iation of the bone caere :, lu ; and CI tract. The total risk is then: R=R 3+ (1-R )R2+3 (1-R g)(1-R )R 2 3 (6.3) where R , R2 , and R 3 are tr.e riskb to the three organs, respectively. MACCS 3 gives sonewnat hignee rise , reircipally because the lung dose is now consid-ered more ef fective 1r. procucina f atalities, and also because the hazard func-tion gives sone risk at lower doses. Tne effect of ite cccc di'fe ences is that MACCS predicts a higher proba-bility of a srall nu .ber c' deaths while CRAC2 predicts a higher probability for large nnters of deatts. Tnis is shown in Figure 6.8 from a corparison calculation perfornec by Sandia National Laboratory for a severe ground level release. Tris is for a u "c r- pc::ulation distribution without evacuation. However, MACCS cea predict substantially more early deaths when evacuaticn is modeled since the lung dose usually becomes dominant, but evacuation scenarios are not considere: i n t rM s re'. i e- . BNL calculated the individual risk of fatalities versus distances for S I 'a' , 52J, and 56W as shoe i h gures 6.2, 6.3, and 6.4 It car. be seer. that the risk is similar to the 200 ren curve, but is not directly correlated because of the nonlinear interactions in the above equations. 6.3 Comparisons of Results 6.3.1 Results of Seatrook Study l BNL used as c basis for comparison the 200 rem risk of whole body dose as i calculated by Seabrook using CRACIT and 200 rem red marrow risk using MACCS at BNL. Tnis is the ris6 to a hypothetical individual located at a particular distance and actual population distribution is not considered. The BNL

6-f calculations we-e re-forre: wit

• neteorological data supplied by Seabroot.

Projected nunbers of f atalities were not computed since BNL didn't have the actual Seabrook po; lation data. Tne Seabrook calculations for accidents SIW, S2W, and S6W are shown in Figures 6.5, 6.6, and 6.7 with the enrresponding BNL results superinposed. For 51W, MACCS predicts slightly higher risk at less than 8 miles and much lower risk beyond 12 miles. The differences may be partly explained by differences in weathe- sa plir.; rethods and plune rise formulations, since this is a high eae gy releasc. Hovtever, the differences in the tail of the risk curve is net corsicerer by ENL to be significant considering the overall uncertainties ir. tre calcoltt;s.. For S2W, the MACCS code again predicts higher risk in close and a some-what sharper dropo'f ir the tail. However, the differerce is that MACCS predicts a risk of .On! at 2 riles wnereas the Seabrook results show this risk at 2.5 niles. For 56W, tFe PNL results are close to those of Seabrook as in the case of 52W. MACCS predicts .031 risk at a niles whereas Seabrook predicts this risk of 200 ren at 6 niles. In su nary, BNL feels tnat tne consequence modeling is fairly presented by Seabrook and that the relatively small differences computed by BNL are probably explaine? by rod +'i t echricues. 6.4 Sensitivity Studies Two categories of sensitivity studies have been performed as part of the BNL review. First, sensitivity calculations were perforned to assess the affect of the duration of fission product release on the dose vs distance curves presented in PLG-0465. These sensitivity calculations are discussed in Section 6.4.1. Second, the impact on the dose vs distance and risk estinates in PLG-0465 of the va'ious concerns raised by the BNL review was assessed in each section of tnis technical evaluation report. These revised risk esti. nates are sunnarized in Section F. 4.2.

S & 6-7 C.c.' Sensitivity of Results to M;1tipuff Release BML performed sensitivity calculations with regard to the raltipuf' releases anc the release duration for 52W and 56W. The results are given in Figures 6.6 and 6.7, respectively using the one puff release categories defined by Seabrook. In both cases it was found that a one puff release increased the risk and also that a shorter duration of the release (0.5 hours) further increased the risk. For the one puff, 0.5 hour release S2W, the 200 ren .001 risk distance increased fron 2 miles to 7 miles anc the 56% distance increased fror 4 niles to 15 riles. This demonstrates that a long release duration leads to a greater plune dilution and less risk at larger distances. Hence, in order tc hase cor.ficence in the Seabrook calculations, one 'must have confidence that the releases will occur with rates and durations similar to those assumed by Seabrook. 6.c.2 Sensitivity o' Results to BNL Review Tne Seabrook study (PLG-0465) identified the following three areas as being the nost influential in calculating the Seabrook risk esticates: The effectiveness of the Seabrook Station prima ry containrent to either remain intact o* to maintain its fissior product retentict capability for periocs much longer than required for even delayed, ad hoc protective actions. A more realistic assessment of the strength and failure modes of the Seabrook containnent than was possible within the state-of-the-art of PRA when the RSS was conpleted. A more realistic treatment of the initiation and progression of inter-facing systems LOCA sequences. 4 Thus, the BNL technical evaluation focused on the areas that were identi-fied in PLG-0465. The approach taken was to perforn sensitivity studies to assess the impact o r. the results in PLG-0465 of the BNL revient of these areas. Tne sensitivity stacies used the conditional risk indd.ces provided in

9 9 , 6-2 PLG-0 W (and suppo-ting docunentation) to assess how changes in the l probability of accident seauences and containment failure modes would change the _ risk estir:ates in PLG-0465. The sensitivity studies calculated revise *d l 200 rem-dose versus distance curves and the mean absolute risk of early l fatalities and total cancer fatalities. The results of the sensitivity  ! studies are given in Table 6.2 and Figure 6.9. The dose vs distance curves in I Figure 6.9 can of course be directly compared with the dose vs distance curves

presented in PLG-0465. However, the mean absolute risk numbers in Table 6.2 1 cannot be directly compared with the information on individual risk provide.

in PLG-0465. The mean absc'. .te tisk number in Table 6.2 would have to be

                                                                                                                         .l converted into individual risi of fatlaities in the population within 1 mile                                   l I

of the site bouncery bef ore direct comparison within information in PLG-0465 could be made. However, the information provided in Table 6.2 is a useful l indication of how the PLG-0465 risk levels in PLG-0465 would change if recalculated using the assurptions of the sensitivity studies. 1 l In addition tc tne areas identified in PLG-0465, other areas were

 'l identified as potentially inportant to risk at Seabrook.          In particular, the                           ,
                                                                                                                            \

applicant was re:;ue sted to provide information- on the risk associated with accidents during shutdown. The results of the applicant's assessment cf such accidents are also given in Table 6.2 and Figures 6.10 and 6.11. BNL was not able to assess the f requeaty of these events (refer to Section 2.2) because there remain fundamental questions regarding modeling of these scenarios. In addition, the potential for induced steam generator tube rupture (SGTR) for accidents in which the prir'acy systen is at high pressure was identifiec ac a  : topic for review. This topic was reviewed in detail by the NRC staff and is I the subject of continuing NRC and industry research activities. In an effort to assess the potential influence of SGTR, a simple sensitivity study was performed at BNL (refer to Section 2.3) and the results are also given in Table 6.2 and Figure 6.9. In Table 6.2 and Figure 6.12 the effect on risk of cordining all the sen-sitivity is presented. This calculation should not be interpreted as a reas- l sessnent of the overall risk profiles for Seabrook. It is simply intended te indicate how the results of the various sensitivity studies could influence the risk estimates in PLG-0465. The method used to combine the effect of all J l

6-9 l 1 of the studies is ret rigorous and could lead to inconsisteacies. In aded - tion, it is not nornal practice in probabilistic risk assessments to conbine bounding sensitivity studies. The results in Table 6.2 and Figure 6.12 should l therefore be recognized for what they are, namely, a series of sensitivity studies and not be interpreted as a statement of the overall risk at Seabrook. The results in Table 6.2 and Figures 6.9-6.12 are useful to focus on those areas of the BNL review that appear to have the greatest impact on the conclusions i r. FLG-0465. Tne results indicate, given the extent of this focused review, that the conservative assumptions regarding accidents during shutdown and induced SGTR have the most impact on the dose vs distance and risk estinates in F 6-O'65. however, the more optimistic assumptions regard-ing these events have minor impact on the PLG-0465 results. These are areas of considerable uncertainty, which at the present time do not allow a better defiritior. cf the risk estir.ates tha*. given by the ranges in Table 6.2 aat Fi gures 6.9-6.12. 6.5 References

1. NUREG-0?9f.

2. RSS.

3. CRA:.

4 CRAC2.

5. CRACIT.

f. MA:05.

7. NUREG/CR-4214

8. BEIR-1.

9. BEIR-3.

10. PLG-0465.

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1 10 1CO 1.C 7. DISTANCE (MILES) . Figure 6,9 Com;ariscr. cf E:.' sensitivity studies with PLG-0465 and frJREG-0396. (200-rem plots with no imediate protective actions)

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6-21 Tarle 6.2 Ir;at: c' :*. Se- " tivi:3 5 u:ies e F.G-T fi F.is. Estinates Absolute Risk Per Reactcr Year Early Fatalities Total Cancers Sensitivity One Two One Two ho Mile Mile No. Mile Mile Study Evac. Evac. Evac. Evac.- Evac. Evac. Original 2.7(-3)* 3.6(-4) 2.4(-5) 1.5(-2) 1.4(-2) 9.2(-3) PLG-0465 Results Reviset 2.F'-2 4.6' 2) 1.2( 4) 1.5(-2) 1.4(-2) 9.3(-3) Frequer:y for Inter-facing Systen LOCAs Steam  :.5.-2; 2.2.- '. 1.8(-3) 1.7(-2) 1.6(-2) 1.1(-2) Generator Tube 2.8l-31 4.2( 4) 8.4(-4) 1.5(-2) 1.4(-2) 9.3(-3) Ruptured Containment 4.7(-3) 2.4(-3) 2.0(-3) 1.8(-2) 1.7(-2) 1.2(-2) Loads and Performance Accidents ** 6.2(-3) 2.9'-?) 3.0( 4) 2.1(-2) 2.0(-2) 1.4(-2) frc, Shut- 2.7(-2' 3.i, 2, 2.R(-5) 1.5(-2) 1. ;-2) 9.2(-3) down**** Inpact ** 1.0(-2) 7.0(-3) 4.2(-3) 2.6(-2) 2.5(-2) '1.9(-2) of all Issues *** 4.9(-3) 2.6(-3) 2.2(-3) 1.8(-2) 1.7(-2) 1.2(-2)

        *2.7(-3) = 2.7x10-3
       **Pessinistic assunptions.
      ***0ptimistic assunptions                                                           ;
     **** Calculated by the Applicant, not confirmed by BNL.

f* %,b

 *'                                       UNITED STATES

[he[$: NUCLEt.R REGUL ATORY COMMISSION gps , v.:s s . a o , c c 2: m yvi j ,

    ~,....'

DEC OB Y PEMORANDUP FOP: Vincent A. Noonan, Project Director Project Directorate f5 Division of PWP Licensing-A FROM: Charles E. Rossi, Assistant Director Division of PWP Licensing-A

SUBJECT:

STEAM GENEFATOR TUBE RUPTURE DURING SEVEPE ACCIDENTS AT SFAcM or STATION - DRAFT INTERIM PEPORT Plant hace: Seabrook Statien, Unit 1 Docket Numbe-: 50 / '. Resp. Directorate: PWR Directorate d5 Project Manager: Victer Nerses Review Branch: Facilities Operations Branch, DPL-A Review Status: Ongoin; Ar accident secuence with the potential to impact risk at pressurized water reactor plants under some conditions is the loss of steam generator tube integrity due to generatic" of high temperatures at hich pressure during a core melt accident. The potential concern involves movement of high tempera-ture fluid from the region of the melting reactor core into the steam cenera-tor tubes with a resultant overheatino of the tubes which leads to steer generator tube rupture (SGTRI. High pressure fluid containing radioactive material from the melting core would thereby be released to the secondarv side of the steen generators, fror where it could be released to the environment via tre stear generator relief valves, thereby bypassing containment. Public Service of New Harpshire (PSNH) has investigated the possibility of encountering conditions in the reactor coolant system under which SGTg can be of concern, and has deten-ined the likelihood to be less than 4 X 10~ per reactor year. This is suf ficier.tly high, and the potential consecuences c' SGTR under severe accident conditions are sufficiently great under some circumstances, that PSNH initiated an investigation of this topic. The enclosed report is a draft copy of an interir document prepared by Warren Lyon (F05-A/DPL-Al which addresses the state of knowledge pertaining to Steam Generator Tube Rupture during postulated severe accidents, and the application of this knowledae to the Seabrook Station nuclear power plant. This report is prepared with the assumption that the work and assessment will continue. The report does not cover all material received from PSNH and its contractors, and PSNH and their contractors have offered to discuss the issue further. It also has not been subjected to comprehensive peer evaluation within the NRC.

Contact:

W. Lyon y280E3 1

l DRAFT  ! Vdree Ar.: :- .:. Tre ra, ice cerclusier.s providet ir tFe dra't rece-t are as follows:

1. Study of SGTR due to severe accident conditions is difficult due to the corple>ity c' the phenorena and the developmental nature of analysis techniques.

2. Further work is necessary to conclude that SGTR is unlikely under conditions associated with a severe accident.

3. SGTR due to severe accident conditions can be shown not to be a problem if the reactor coolant system is depressurized.

L b @ Cherles E. Rossi, Assistant Director Division of PWR Licensino-A

Enclosure:

As stated cc: T. Novak S. Lon;

f. Coffr.ar J. Her t

V. Leunc V. Eenaroya S. hewberry 4

                                                                                   -r l

E f.'CLOSUP.E SEABROOK STATION STEAM GENERATOR TUBE RESPONSE DUPING SEVERE ACCIDEr:TS DECEMBER 5, 1986 DRAFT t I

DRAFT Coreworc This re crt addresses the state of knowledae pertaining to Steam Generator Tube Rupture during postulated severe accidents (approach to core melt art core relti, and the application of this knowledca to the Seabrool Statier nuclear power plant. This is an interim report, prepared with the assurotier, that the work and assessmert will continue. The report does not cover all raterial received fror Public Service of New Hampshire (PSNP) and its cor. trac-tors. It also reoresents the view of the author and has not been subiected to comprehensive peer evaluation within the NRC. 1 .~~ -- - --r -

OP A FT ho~er.clature AC Alterritin; currert ENL Brookhaven National Laboratory EPel Electric Power Research Institute ICC Inadeouate core cooling KK Kilowatt LM Larson Piller parameter LOCt Less of coolant accidert Fa' Meoewatt NRC huclear Regulatory Connission NSSS Nuclear steam supply system PDS Plant damage state (See belowl

       . POPY      Pressure operated relief valve PRA       Probabilistic Risk Assessment PSt#      Public Service of New Hampshire RAI       Request for additional information RCP       Deacter coolt't pu ;

RCS Reactor coolant syster RWST Refueling water storage tank SG Steam generator SGTR Steam generator tube rupture Plant damage states are used to classify conditions as follows: 1 Early core melt, low RCS pressure at tire of reactor vessel failure, PWF' injection not initiated 2 Early core melt, low RCS pressure at tire of reactor vessel failure, PWST injection initiated 3 Early core melt, high RCS pressure at time of reactor vessel feilure, RWST injection not initiated 4 Early core melt, high RCS pressure at time of reactor vessel failure, PWST injection initiated E Late core melt, low RCS pressure at tine of reactor vessel failure, RWST injection nct initiated 6 Late core melt, low PCS pressure at time of reactor vessel failure, PWST injectier initiated 7 Late core melt, high RCS pressure at time of reactor vessel failure, PWST injrction net initiated S Late core melt, high RCS pressure at tire of reactor vessel failure, RWST injection initiated 9 Core melt with non-isolated SGTP 2 A Containment intact at start of core melt, containment heat and fissier product removal available 3

DRA FT F Cc.; tin-art init:- 2: start c' core relt, cortainmer.t heat reroval oriv astilatie C Containmer.t intact at start o' core melt, containment fission product rereval only available D Containment intact at start of core melt, none of the containment +urc-tions available E Containment not intact a+ 5 tert of core melt, activity release filtered F Containment not intact at start of core melt, containment opening larger than three inch dianeter FP Containment not intact at start of core melt, containment openina smaller than three ir.ch diametcr FA Aircraft crash ( e J 4

DPAFT

    ;, cyEr :E, t*c sy m The Public Service of New Hampshire (PSNHI has presented informatier. to show that the Seabrook Station containnert is one of the strongest of any nuclear power plant. It also contains one of the largest volumes. This combinatirn leads to a conclusion that the containment has the capability to either sionificantly deley er neavant the release of large cuantities of radioactive material during and following a severe (core damage or core melt) accident.

Based on this premise, any significant risk associated with Seabrook Station would likely be found in accidents which bypass containment. A number of potertial bypass possibilities exist, some of which have tradi-tionally been recognized ir. Frobabilistic Safety Assessments (PRAsl, and some of which have not. Historically, conservative assumptions have been applied to those cases which have been recoonized, and the conservatism has been assumed to be sufficiently large that the unrecognized possibilities becane insignificant since they were believed small in comparison. PSNH and its contractors have provided a comprehensive PRA with additional follow up investioations in which a better representation of nuclear plant behavior has been attempted. Some conservative, and thereby misleading, representations have been removed. This approach to accident analysis lear's to the possibility that when conservatisms have been removed, previously neglected bypass kaths which were masked may now be found to contribute te risk. Recognizing this, the Staff and PSNH have explored containment bypass possibilities. One pessibiliu , the topic of this report, and a poter.tial issue that has been under investigation by industry and the Staff for several years, is the loss of steam generator tube integrity due to generation of hich temperatures at high pressure during a core melt accident. The potential concern involves movement of high temperature fluid from the region of the melting reactor core into the steam generator tubes, with a resultant over-heating of the tubes which leads to their rupture. High pressure fluid containing radioactive material from the meltine core would thereby be released to the secondary side of the steam generators, from where it could be released to the environment vie the stean gererator relief valves, thereby bypassing containment. 5

i l l DPAFT l Fer 5 Ea" cere-air- tuu rut:ure (SGT;' tc be a conce-r. as addressed here, ort rust have a cere darace (cr relt i condition in progress with ro water e- the steam generator secondary side. The principal contributor to this conditier. is estimated to be a loss of all AC power concurrent with a loss of all turbine driven feedwater to the steam generators. PSNH has investigated the possibility of encountering cont'itions which can contribute to SGTR and has determined the likelihood to be less ther 4 X 10-5 per reactor year. This is sufficiently hich, and the potential consequences of SGTR under severe acci-dent conditions are sufficiently great, that further investigation has beer necessary. This investigation is ongoine. This report provides an interim assessment of the status of the investigation, as well as a projection of expected results. Study of SGTR due to severe accident conditions is difficult. The phenomena are complex, and most analysis technicues used to investicate nuclear power plart behavior have utilized assumptions which are not applicable here. The principal complication is the multidimensional character of fluid behavior ir. the reactor coolant system. Suitable computer programs are ,iust beginning to becere available. Suitable experimental information is just being developed. Hence, pioneerir.g work, such as provided by PSNH in investigatior. cf this issue, car. be expected tc have weaknesses as well as strengths. We have found this expectation to be true. The work reported by PSNH and its contractors is highly infomative and addresses most aspects of the SGTP issue. It is based upon knowledge of what takes place within the Nuclear Stear. Supply System (NSSS1, upon a major computer program that is under development and is being verified (MAAP), and upon infortr.ation derived from an experimental program at Westinghouse. The following is a sumary of the reported infortnation and our assessment:

1. Pathematical modeline. Expected phenomena, experimental information pertinent to the phenomena, and modeling assumptions have been addressed for each of the major components of the NS.SS which are affected. Multi-l dimensional fluid flow and enerov transport have been established as ., I dominant over most of the conditions of interest. We consider this area l

6 l

DRA FT te is ir e r +1irir.a ry stegt o' development, anc there art scre Detential difficulties, which include:

a. Certain modeling assur.ptions are overly optimistic. An example is the assumption of complete mixing in the steam generator inlet plenum which tends to reduce the temperature o' fluid entering the steam generator tubes. This assumption is not supported by the available experimental evidence, and the effects of the assunption are not balanced by identifiable pessimistic assumptions elsewhere in the analysis.

b. Experimental evidence is preliminary. The experimental facility at Westinghouse is providing information pertinent to this issue.

However, testing has been limited to conditions which are only roughly scaled to NSSS representation. This is due to a logical progression in the test planning and facility development. Da ta fror apparently well scaled test conditions are just becoring available. he other test facility addresses certain aspects of this issue.

c. The co ruter prograr used as the basis for much of the work has net been verified, nor is docurentation available. We understand a verification prograr and an effort to provide documentation are unde rway. (PSNH contractors have offered to discuss this informa-tion with us. Our review has not progressed to the stage where we car. r. ale use this offer.) Although the phenomena we understand to be modeled by the code appear adequate for the purposes needed here, and the code results appear reasonable subject to our concerns as expressed elsewhere in this report, this is not sufficient information to accept the analysis results.

2. Seabrook Station Peoresentation. The basic enalyses and sensitivity studies have been based upon a plant configuration in which the NSSS state is assumed. Most of the assumed state conditions are reasonable.

There are exceptions. For example, the steam generator secondary side is 7

DRAFT assu ti tc te a a tressure cceresponding to secondary side relief velvt settings, and creer rupture of tubes is reported for this state. The resulting conclusions are similarly based upon this state. We believe there is sufficient likelihood the secondary side will be depressurized that this case should be considered. Depressurization would roughly double tube stress since the secondary side pressure would be decreased from roughly 1100 psi to atmospheric pressure while the RCS pressure remained at approximately 2300 psi.

3. Sensitivity Studies. PSNH and its contractors have performed a wide ranging sensitivity study as part of an assessment of the impact of various modeling assumptions and the state of the plant. Although this yields valuable information and insight, sensitivity studies should be approached with caution. They are only as good as the basic modeling.

The impact of our difficulty with assumptions such as the behavior of the stear generater initt plenum is not addressed in the sensitivity study, and could impact the results and conclusions. 4 Operator Actions. Plant response can be drastically altered by operator actions during a severe accident. SGTR is no exceptier,. A number of operator responses have beer discussed with PSNH. Although many cf these were postulated actions, sienificant information has been developed fro-these postulatiers. Fecognition that operator actions could depressurize - the steam generator secondary side is one item raised during the review. Depressurization of the reactor coolant system via the pressurizer 1 Pressure Operated Relief Valve (P0FV1 to avoid the SGTR problet is l another. 1 Ve find that the topic of SGTR is in a developing state, with knowledge being rapidly accumulated. Further work is necessary to conclude that SGTR is unlikely under all conditions associated with a severe accident. Existing knowledge can be used to support a conclusion that SGTR is not a problem if the RCS is depressurized. Consequently, reasonable assurance that progressions toward core melt would not occur at high RCS pressure, coupled 8

i DpArT l with ser :*tir: etici te ir. recard te steer generater tube respensE, WOU f l i alleviate our concern recarcing SGTR under severe accident conditions. Ve have net conducted an evaluation of the trade-offs associated with such an approach, ner have we been provided with information that would either suptcrt or negate RCS depressurizatier. Under severe accident conditions. Ve have not provided a recommendation regarding whether RCS depressurization is attractive when all pertinent factors are considered. Our judgement is that a carefully conducted thorough evaluation on the part of PSNH can establish that the likelihood that a SGTR will result due to overheatine during severe accidents which initiate from power operation is sufficiently small that the risk associated with this event can be shown to be negligible. Our judgement is prelirinery and has not been substantiated. Substantiation of a judgement regarding SGTR under severe accident conditions originating from pcwer operation with the RCS at high pressure can be obtained through a conbination of analytic and experimental investigations. The engring test at Westinghouse in which reasonably close similitude is claimed between the test facility and appropriate parts of a Westinghouse four loop NSSS will provide key data which can be applied to assist in the development and confirmation of analysis techniques. Use of selected test data from other facilities and further examination of the analysis techniques, coupled with necessary changes when they are uncovered, should provide sufficient confirra-tion that reasonable reliance can be placed upon accident analyses pertinent to this issue. Suitable analyses can then provide a sufficient foundation to resolve this issue. 9 l l l

DDAFT

2. INTRO M Tire The Public Service of hew Hampshire (PSNH) reporting of Seabrook resperse to accident conditions in References 1 - 4 represents one of the most comprehen-sive investigations of nuclear power plant accidents in a specific plant that we have encounteced. Sore accidents which have a significant impact upon risk are treated more comprehensively than previously reported by any investigator.

For example, References 3 and 4 describe an investigation of LOCA outside of containment that is more comprehensive than any we have reviewed. Many of the ~ comenly used conservatism, wHch distort the perception of accident impact, have been removed. What results is a serious attempt to better represent plant response tc severc c::ide-t conditions, with particular attention to items which have previously been identified as having a serious impact upon risk. Paradoxically, as will be seen, this attempt to better represent plar.t behavior requires a core careful review of certain aspects of severe accidents than recuired for previously reported PRA investigations. PSNH has presented information to show that Seabrook Station has one of the strongest containments of ary nuclear power plant. It is also one of the largest with respect to containment volume. The combination of large volure and strength leads PSNH to a conclusien that the containment can mitigate virtually every severe accident and, at the worst, can significantly delay release of meaningful cuantities of radioactive material during and following core melt accidents. Most core melt accidents crn be contained within the Seabrock Statien containrer; ard, if this is accomplished, little radioactive material will escape. The full mitigative capability of the Seabrook contain-ment will be realized if there are no

• holes" in the containment. Such holes i can exist if any c' the following occur:  !

1. Containment is not properly closed fisolated), such as can occur if containment ventilation is not properly closed upon receipt of a contain-ment isolation signal,

7. A failure occurs which allows the containment atmosphere to escape, such as failure of a containrent penetration due to a corrbination of high pressure and high tenperature, or 10

DRAFT 3. A 'ailure occuer whic' allws caterial to ecve cirect'y fre- the f.aclea-Stea- Sur:1) Syster It.SSS',, principally the Reactor Coolant System (RCSi, to the environment, such as occurs with the traditional " Event V" (Pef. 5), with LOCA outside conteinment leading to core nelt and the release of radioactive material via the LOCA flow pathway. Clearly, if PStai conclusions regarding containment strength are verified, there will be little risk associated with accidents at Seabrook Station unless containment is bypassed. Therefore, core damage accidents with containment bypass deserve careful attention. PSNH has reported studying some bypass accidents in significant detail (Pefs. 3, 4, I?,17, and 18). Such studies have led then to conclude that certain bypass accidents at Seabrook, such as LCC outside conteinre"c. engender significantly less risk than previously believed. In PRA investigations, one iray neglect some small contributors to risk since they are negligible ir, comparison to major contributors. If ma,ior contributors i are found to be significar.tly smaller, then one must check the previously neglected contributors to assure they are still negligible or, conversely, they must be included in the contribution to risk if they are now significant. This is the situation that is typified by the Seabrook Station PPA investigation (Refs. 1-4). A major contributor to risk associated with containment bypass, Event "V", has been analyzed by PSt?, and they have concluded that it is not the significant contributor to risk that it was previously believed to be. This situation has led us to ask "Are there any containment bypass risk contributors which have been missed or which require further consideration?" l One potential area for bypass, as identified above, involves a path between the RCS and the environnent. One way of searching for such paths is to ask "Are there any phenomena which may occur, and which have not been adecuately addressed in past searches for accident possibilities?" We and PSNH, among others, have asked that question, and found that certain phenomena have been neglected in past PRA investigations because their contribution to risk is small in comparison to other contributors. The phenomena of potential signi-ficance involve multidimensional fluid behavior and fission product transport 11

, . o DRAFT , vittir tr4 *:S c *it; t*e acpecact tc core trelt and durinc the core mC t i ' orocess. Consideration of these phenomena has a sicrificant impact upon PCf response, including potentially the location of RCS' failure. There are many possible irplicatiers, including the possibility that the impact of RCS failure on containment may have been overestimated in past analyses. The implication of interest here is that failure to accurately model RCS fluid and fission , product heating behavior might result in an RCS failure which bypasses con-tainment. The only area discovered where this is of 1 mediate concern involves the Steam Generator (SG) tubes. If these fail during a core melt accident while the RCS is at high pressure, there is a high potential of a major release via the SG relief valves or the SG Pressure Operated Pelief Valves (PODVs), which vent directly to the environment. The general concern addressed in this report is the rupture of SG multiple tubes in response to high temperature, which in turn is a result of core uncovery. This accidert secuence should be of concern any time there is a core melt with the PCS at high pressure in combination with no water in the SG secondary sides. These conditions lead to a potential for natural circulation transport phenorena to significantly heat the tubes prior to breach of the reacter vessel. If this occurs, the resulting loss of tube strength could lead to tube rupture. If tube rupture occurs, and any of the secondary side valves are open, the secondary side is breached outside containment. Al te rne-tively, if the RCS pressure is above the SG relief valve setpoints, contain-ment is similarly bypassed. This has not been adequately investigated, and is not recognized as a release path in the early Pickard, Lowe and Garrick work or risk investigation at Seabrook Station (Refs. 1 4). It has been addressed in more recent work (Refs. 12, 17, and 18). The concern was expressed as the rupture of multiple steam generator tubes. We do not believe single tube ruptures will occur under the severe accident conditions of interest. The reason for this is that if one tube ruptures, nr even begins to leak significantly, this will indyce flow of hot RCS fluid toward the leak. Therefore, the location of tube rupture will probably quickly become hotter. If high temperature is what led to the break, a higher temperature can only make it worse. Tubes in the vicinity of the break will

,      be exposed to the high velocity break flow, in additional to high temperature, I?

l _ - . - . ~ _ _ . - . _ ._. _ _ - - - .

I i DDAFT l weakening the e*f. we beliese, ouickly leading to their failurt. ke believe this cascading effect would rapidly propagate to multiple tube rupture, stopping only when sufficient RCS depressurizatinn has occurred that tubes are no longer stressed by a significant pressure differential across their walls. Although this report is limited to SG tube rupture, there are other SG corpc-nents which separate PCS fluid from the SG secondary side. These components, such as the SG tube sheet, must be investigated to achieve completeness in the investigation of containment bypass via the steam generator. An initial consideration in investigation of the SG tube rupttre issue is "What is the likelihood of attaining conditions where SG tube response could be of concern?" Principally, the conditions are loss of all SG feedwater witt

       ca simultaneous loss of RCS makeup capability; conditions which result, for example, from a loss of all AC electrical power with the simultaneous loss of the turbine driven auxiliary feedwater punp. PSNH estimated this conditinr. tc have a rear annual frecuency of d.5 X 10-5 per reactor year (Pef. 17), a value sufficiently high that tube response rust be considered.

13 1

         - - + + -     y       w--                -     ,y      -     w-               --

w

e, . DCAFT y I STEtv sp rr r: "er reinrE f SSTF) UNSEF SEGE ACCICEN~ C0'2:' MI l 3.1. Description of Phenonena and Potential Concern. The RCS is generally nodeled with a one dimensional representation of fluid flow, and in some cases with parallel one dimensional modeling in regions such as the reactor vessel. This has been particularly true for PRAs, where to our knowledge, all have been based upon cornputer code analyses which incorporated single dimensional representations of fluid behavior within the PCS. Addi-tionally, movement of the source of heat due to fission product r'igratien is seldom modeled. The possibility of RCS behavier being different from what is generally repre-sented during severe accidents has been recognized for some time. Winters (Ref. 6) identified aspects of the problen in 1982 and Denny and Sehgal (Ref.

7) provided preliminary multidimensional analysis results in 1923. It was the subject of an NRC/ Industry meeting (Ref. 8), and a formal request for work within NCC (Ref. 9), in 1984 Potential impact upon SGTR was estimated on a preliminary basis (Ref.10), and experimental data were presented fror an ongoing series of tests (Ref. 11), in 1985. Numerous analysis results have been published since the early publications of Denny and Sehgal which repre-sent work sponsored by both industry and the NRC. However, there is no published analysis of overall NSSS response to a broad range of severe acci-dent conditions which includes these phenomena, and which is based upon accident analysis methods which have been subjected to broad peer review and acceotance. This introduces a difficulty into review of SGTP during' severe accidents with respect to the impact upon the Seabrook Station risk evalua-tion. As will be seen, sufficient work has been accomplished that what appear to be reasonable conclusions can be formulated, although confirmation will require additional effort. As will further be seen, there appear to be operational methods which can negate the problem, althouoh the impact on other aspects of plant operation has not been evaluated.

The one dimensional analysis approach appears adequate for approximation of { the time to core uncovery following accident initiation. Whether it is adequate tn represent behavior 'nllowing core uncovery depends nn a number of 1t 1

                                                                                       - =

l

a. .

1 DRAFT

      #acter:, su:& et th cerse*,Etise associated witF PCS mccclin;. syste re-spctse te accident ccrditions, and the type of accident. 1r general, one should cuestion the adequacy if two, and sometimes one, e' the followir.g conditiers exists:

1. RCS pressure is in the vicinity of nomal operating pressure

2. A liquid " plug" exists at RCS low points (the lower reactor vessel or the crossover legs between the SG exit and the Reactor Coolant Pump (RCP) suction connection), and the remainder of the PCS is filled with vapor er gas

3. Accident contributors cererally found to be major contributors to risk have a lower probability of occurrence than found in most PRA investigations in the case of Seabrook Station, all three conditions apply with respect to SG response. A number of conditions potentially lead to core melt with the FCS at high pressure (including, for example, a loss of all AC electrical pover with loss of feedwater), such conditions are calculated to leave liquid plugs in RCS low points for Westinghouse designed NSSS's, and the Seabrook Station PPA work represents corprehensive modeling with removal of some unrealistic conservatisms.

The potential misrepresentation of system response of concern here stems frer the fluid flow behavior inherent in one dimensional modeling. Such modeling typically represents flow through the reactor core as detemined by the water boiloff rate from the lower core or lower plenum. This rate becomes small as the water level approaches the bottom of the core. Typical calculations (see , historical references which were previously discussed) indicate ~that the flov I rate due to natural convection which occurs in a multidimensional manner is of i the order of ten or 'nore times that of the flow due to boiloff. Hence, the calculations are typically based on a minor contributor to. flow, and the major contributor is neglected. The modeling difficulty also applies to upper plenum behavior. One dimension-al modeling of any fluid (licuid, vapor, or gas) that passes through the core is typically assured to flow throuch the upper plenum and out the hot leg. 15

a. .

DPAFT

       'Mt redt' ire is it: tere:t urder severe accidert conditions wk.ere a et de-pcrtior. of the core has beer uncovered or the core is being vapor or gas cooled since strong recirculation patterns will develop which therrally lini the core and urper plenum. At pressures in the range of 2250 psi, the linkage is strong, and sore of the upper plenum component temperatures can be expected to closely follow core temperature during the early stages of the approach to core melt. The strength of the linkage diminishes with decreasing pressure.

Information also exists which illustrates a decrease in linkaoe with increas-ing hydrogen concentration and core damage (although initial production of hydrogen may enhance circulation due to the buoyant gas " pushing" its way toward upper regions of the reactor vessel). Similarly, correct entsider?tice o' the hot leg and steam generator behavior leads to calculation of significantly different behavior when contrasted to one dimensional modeling. Hot fluid, at a temperature far greater than predicted via a one dimensional rodel, will enter the upper portion of the hot legs fror the reactor vessel, and flow toward the inlet plenum of the steam generators. Displaced ccider fluid will return to the reactor vessel upper plenut along the bottor of the hot legs. Circulatory patterns will become established in the steam generator inlet plena in which some of the hot incoming fluid is mixed with plenur fluid. Fluid from the steam generator inlet plena will flow into some of the steam generator tubes in the nominal forward direction, displacing fluid in the stear generator outlet plena. This displaced fluid will flow through other tubes in a nominal reverse direction, reentering the steam generator inlet plena. (All of these flows have been observed.experi-rentally as described in Peferences 11, 13, and 14). This mechanism has tFe potential to transport hot fluid from the reactor vessel into the steam generator tubes durino core heatup and melt, with the result of creatino the potential of overheating the tubes if there is no water on the steam generator secondary side. l There are other possibilities which could challence tube integrity as well. { For example, many plant Inadeouate Core Cooling (ICC) emergency procedures  ! specify RCP operation if conditions exist which indicate an approach to core melt, and alternate mitigative measures have failed. Such a step could circulate het fluid through the RCS, including the tubes. Although this may 16 1

DRAFT slightly extert it{ titt 1 core rO t, it may be at unattractive approach if it also introduces a high likelihood of loss of tube integrity. To our knowledge, these contrasting responses and the impact upon risk have not been studied. (Note the likelihood of encounterirg the situation is small.) A final phenomenon that has received inadeouate attention during conditions leading to core melt is fission oroduct movement. Typical one dimension accident code calculations take such movement into account from the viewpoint of radiological hazard, but do not include the influence upon heat generation. Approximately a quarter of the heat producing radioisotopes probably has lef t the core under the corditiers of interest, and substantial deposits can be expected in the upper plenum structure. This could have a significant influ-ence upon thereal respor.sc, perticularly if some of this material leaves the reactor vessel and enters the hot legs. As will be seen in the following sections, PSNH has addressed many of these issues in the most corrprehensive study of this probler that we have encoun-tered. 3.2. Seabrook Statier Steam Generator Integrity 3.2.1 Issues Aodressed By PSNH The PSNH has addressed many of the issues applicable to SG tube response to severe accident conditions (Refs.12,17, and 18). Analysis results were su rarized which wert intended to deterrrine the therral response c' SG tubes under severe accident conditions. Basic analysis assumptions pertinent to the state of the plant were:

1. The steam generators must be dry to experience a significant thermal transient since, if the SG secondary side contains water, the tubes cannot overheat.

2. Station blackout conditions (Loss of all AC power) exist.

Analyses were conducted for the following: 17

i l I DRAFT

     .. Stati:- bit ti c;t witt.a. crerator actions or FCr seal LCG i

2. Staticn blackout with a 50 gpm RCP LOCA (each RCP) and no operator actions

3. Station t'lackout with operator actions

4. Uncertainty evaluation Possible operator actions considered included:

1

1. Start steam turbire driven auxiliary feed water flow

2. Restore emergency AC power (diesels and/or switchgear)

3. Shed nonessential loads 4 Open RCS PORVs when core exit temperatures exceed IP000 F.

A nurter of other operator actions one r.ight expect were discussed during a meeting with the PSNH at BNL on October 17, 1986, includine:

5. SG blowdown and depressurization to enable filling the SGs by the condensate booster pumps or from fire water systems. (There are two l diesel driven pumps and one electrically driven pump at Seabrook l Station. The ability to use these for injection into the SGs has not been confirmed.)

6. RCP operation, a step that is not possible unless off site electri-l cal power has been restored. (PSNH felt the likelihood was suffi-  !

ciently low that there would be neoligible effect on risk.) l l 3.2.? Likelihood of Conditions Leading to Tube Failure l i PSNW addressed the ouestion of conditiers necessary for SGTR in the i response to the Staff Recuest for Additional Information (RA!) 47 (Pe'. l 18 i l i I i l l

DRAFT l. Ir t! rester.sc , est- s ta ted the vie te te sta li for tr e follo. . ing reascr.s:

1. The frequency of high pressure core melt with dry steam generators is very small.

2. Given the postulated occurrence of a high pressure core melt with dry steam generators, creep rupture of the SG tubes is not a credi-ble failure mode.

3. A large number of tubes must fail to produce an early large contain-ment bypass.

4 All three of the following must occur in order for there to be a containment bypass:

a. Failure to recover water to the SG

b. Failure to decressurize the RCS

c. SG tube creep failure 3.2.3 PORV Considerations POEV operation as identified in item 4, above, is not specifically contained in Seabrook Station emergency procedures, but is believed by PSNH to be a logical operator response as an attempt to depressurire and obtain water from the accumulaters. (Operator monitoring of the terpere-tures is specifically identified in the procedures for loss of all AC powerconditions.) in addition to potential core cooling via the accumu-later water, opening the PORVs is claimed to have the following effects:

1. It reduces stresses in all primary system components 4

7. PORY flu overrides natural circulation such that high fluid temper-atures are not attained in the SGs, including the tubes.

l 19

DPAFT Ir restern it a sis" ovestior., PSn indicatte that the likelihoce c e being able to open the DORVs under loss of AC and ICC conditiers was Hgk (See Secticn 2.2.9). They also indicated that one PORY was sufficient since its " worth" is about 50 W of energy removal in the form of stear, and have presented blowdown rate infortnation in Reference 18. (hote Seabrook is equipped with two P0pVs.) Although we consider the EPRI funded Westinghouse tests pertinent to this issue to be somewhat preliminary with respect to scaling to N$55 condi-tions, some interesting effects have been observed that are worth noting which pertain to PODY operation. These include:

1. hatural circulation flow restores itself readily to the pre-opening condition in the hot legs, core, and connunication paths between the upper plenum and the upper head following PORV closure.

2. Heat transfer in steam generators between the primary and secondary side fluids increases 50% to 75t with periodic venting.

3. The core is little affected except for the boundary with the hot let that connects to the pressurizer surge line.

Item 2 is of particular interest since it carries an implication that flow in the steam generatar tubes is enhanced by PORV operation (as well as by opening and closing of RCS safety valves). Hence, if one visualir-es coening and closing a pressurizer PORY when degraded conditions are well established with the steam generator secondary side depressurized, j there r.ay be a tendency to enhance flow of hot RCS fluid throuch the l tubes, with the pntential of causing tube rupture. 3.2.4 Loop Seals l l Loss of RCS inventory under natural circulation conditions (PCPs not running) is expected to leave the PCS in a condition where water is trapped at low elevations. According to a number of preliminary analy-ses, these exist at the cross ever leg between the SG exit and the RCr 20 1

b

• DRAFT inlet, a*f it t4 lon t region cf the reactor pressure vesse'. The abserte of these water seals could significantly change circulatcry conditions during ICC conditions, with the potential for changing SG tube response. Although we expect a careful examination of behavior in the Seabrook RCS would establish that the seals will remain under most boil down conditions, this expectation needs to be substantiated by suitable analyses which address the ranga of conditions which can exist durina severe accidents.

Complete loss of the RCS licuid inventory with the RCPs running, followed by loss of the RCPs, could result in a homogeneous fluid condition in the RCS. Under this condition, fluid heated in the core would flow into the upper plenut, thrcugh the hot legs, the steam generators, the RCPs, and back into the reactor vessel and the core via the cold legs. Although multidimensional fluid flow conditions probably exist in the reactor vessel after RCPs are lost, one n:ay estinate that themal response is still reasonably realistic if modeling is restricted to one dimension provided the natural convection flow rates are high. For this case, existing analysis codes could be applied to roughly estimate stear generator tube response. If the response was not clear, then multidimen-sional analyses could be applied to estirate the influence. In such a case, uncertainty in the multidimensional analyses mioht not be of as greet a concern as for the situation of multidimensional behavior dorri-

,                              natine systen response. However, nonexistence of the loop seal due to continuous RCP operation is an unlikely situation since the ma.4ority of conditiers during which stear generator tube integrity is of concern will involve loss of off site AC power, and RCPs will be unavailable. To our knowledge, a complete, accurate, analysis of a four loop Westinghouse NSSS has not been performed for these conditions. In edditional to an analysis approach, closure of consideration of this aspect of SG tube behavior could be obtained if the probability of occurrence of the RCS homogeneous fluid condition was established as negligibly small in contrast to other situations where SG tubes were shown to lose integrity, or if the risk associated with the condition was established as negligi-ble when compare / to other Seabrook Station risks.

21 ' l 1

DPAFT 4 sece-f si w ic- ire. O ir- free circulatier in the RCS might be et-tairec if one censicers the RCPs as being resterted in response to higt core temperatures, as prescribed in the emergency procedures. For this case, sufficient head might be developed to clear the loop seals of water, and rehomogenize the PCS fluid, thereby generating the condition described in the previous paragraph. To our knowledge, rehomogenizatier under these conditions has not been established to occur at Seabrook. Insofar as SGTR at Seabrook is concerned, the issue can be dealt with as outlined in the previous paragraph. A third situation of rer'ovel of loop seals also potentially exists during boil down of the RCS inventory. One may postulate that the ICC condition occurs with the leep s u h in place, and that some other mechanism causes their disruption. This could occur if a sufficient pressure difference occurred across the seals that they were forced out of the low regions. Several analyses have been conducted which include consideratien of this behavior, and none showed loss of the seals. One would expect that consideration o' this condition could be closed if analyses applicable to Seabrook could reasonably establish that the seals remain. A final conditien can be visualized if one considers a LOCA to have occurred in the RCS. For example, a small cold leg LOCA for an RCP seal LOCA) could be located between the two natural seal regions of the crossover leg and the reactor vessel lower planum. Removal of RCS mass might occur under conditions such that the seal water was forced out of the RCS via the break. Elirination of consideration of this effect with respect to impact upon risk could be considered on the basis of a themal-hydraulic investigation of RCS behavior, establishing that the potential impact on risk of the behavior is negligible in comparison to other established risk contributors, or both. 3.2.5 PSNH Modeling Considerations The PSNH has reported application of the MAAp 3.0 code to investigation of natural circulation flow in Seabrook (Refs.12 and 17). This code treats the me.ior phenomena, including approximations of multidimensional 22 \

DRAFT flow and fissicr prci.:t (htting) movement, and is applied tc th regions o' the RCS which are affected by the SGTP issue. Quasi-steady momentur balances and continuity equations are used to represent r.atural circulatior, flow, and the steam generator inlet plenur. behavior is represented by quasi-steady inixing models. The modeling represents gas and wall temperatures using conventional lumped parameter models, with 15 gas control volumes and 17 two dimensional heat sinks. (Several volumes are subdivided into further volumes for some types of calculations. The core, for exariple, contains 70 nodes which comprise the core volume node.) The control volurnes are l,ased upon approximations of the flow patterns which were seen in the Westingbouse experiments on a scaled NSSS (Refs. II, li, and 14). This basis for definition of control volumes means that deviations from the assumed flow pattern and flow instabilities may not be represented in the model. Experimental evidence shows that there are asyrretric flow patterns, for example, which are not modeled, and which could lead to tube heating conditions which would not be calculated. Further, elthough instabilities have not been experir:en-tally observed at the Westinghouse test facility, one must accept this evidence with care since testing with fluid conditions which closely simulate those expected ir at NSSS are just being initiated. Use of the lumped parameter model requires further discussion. Unlike co puter codes such as COWlX, which can deter,nine flow patterns within certain bounds provided the configuration is properly modeled, a lumped parameter rodel is based more strongly upor a presupposed flow behavior. Although such representation can be valuable and accurate under certair, conditions, such assumed behavior must be verified before it can be accepted. The preliminary Westinghouse experiments, as discussed brie'1y in the next section of this report, and some COMMIX and MELPROG calcula-tions (Refs. 15 and 16), represent steps in this direction, but further evidence is necessary before we can accept the assumption as verified. 1 (The experiments are somewhat preliminary, and the COMMIX and MELPROG l calculations have not, to our knowledge, been carefully checked against experimental evidence.) We further note that, to our knowledge, there has been no independent study of the version of the MAAP code used for t P3

DRArT tFE ana ly s t : . It a rir.irur, we believe a reasenat.it kr.ciled;t cf codt modeline and logic, ir, addition to a verification prooran, are necesserv for acceptance cf the calculated results. (We note that EPRI has a FAAP verification program underway.) One aspect of the modeling appears worthy of further consideration. The steam generator inlet and outlet plena are assumed to be completely mixed in the PSNH studies being reviewed here, and they are represented by single nodes with uniform properties. The Westinghouse facility test data indicate a partially stratified, partially mixed SG inlet plenur (Ref. 14), and modeling for the test facility is based upon a quasi-steady state model in which partial mixing is assumed at various (limited) locations between streams of different origins. Reference 14 describes the situation as follows:

                *The flow in from the hot leo rises rapidly in a plume in the inlet plenue and induces mixing. Some of the cold return flow from the tube bundle does avoid rixing, particularly near the divider which is furthest from the hot leg. Much of the cold return tubes' flow plunges through the hotter stratified fluid layer that spreads across the bottom of the tube sheet. The mixing flows could be observed from dye injection and from observation of light through the density gradients that resulted. Temperature measurements in the inlet plenum are indicative of mixing. The tubes carrying hot fluid from the inlet plenum were generally concentrated in the area above the hot leg entrance and scattered in the regions further away. Cold return tubes were also scattered and were found in the area above the hot leg inlet also."

Test facility modeling of the phenomena uses a six ecuation approximation which contains an experimentally determined mixing parameter. We believe the assumption of complete mixino used for the PSNH investice-tions will reduce SG tube temperatures when contrasted to the experimen-tally identified situation. This modeling and its implications need further consideration. (This comrent is repeated a number of times in

                                                          ?4 i

D

• DRAFT the discussier c' calculated N555 response in the fcilowine sectic.ns c' this report.)

3.2.6 Comparisons of Calculations to Experinental Date Several comparisons between MAAP code calculations and experimental data have been briefly described by PSNH and its contractors to the BNL and FRC staffs (Refs. 12, 17, and 18). These are discussed below.

1. Core and uoeer plenur flow rates. The following comparison of experimental and calculated values was presented:

Test Condition Experimental Calculated Flow Rate Flow Rate 28 KW Water Test 0.5a 0.50 0.9 KV 5F6 Test 0.016 0.017

2. hot leg and steam cenerator natural circulation. Compariser cf several paraneters was provided:

Calculated Values for Indicated humber of Steam Generator Tubes Experimental Carrying Flow in the Out Direction Iten Value 6 12 24 Heat Transfer Pate, KW 2.43 2.0 2.6 2.9  ! Entering Fluid, CC 30 30.7 29.2 28.4 Exiting Fluid, OC 19 24.2 21.7 18.P. Coolant, OC 10 - 11 9.4 11.2 12.2

                                                                                     )

where the entes fng fluid is flowing into the steam cenerator inlet plenur. from the upper portion of the simulated hot leg, and the exiting fluid is l flowing from the lower portion of +he steam generator inlet plenum back toward the sir.ulated reactor vessel along the bottom of the hot leg. The 25

1 DRAFT cochrt te. rernure is tr.M c' the water leaving the secondary sice of the simulated steam generator, and thus, can be related to tFe heet transfer rate from the primary to the secondary sides. These results are clearly promising. Continuation of the comparisons with a wide range of experimental ccnditions in the same test facility, and with no changes in the modeling except for the change of experimental conditions and fluid properties, would be helpful in code verification. Extension of the same modeling approach to other experimental data (such as flow in ducts and components) would provide further confi mation. Completion of confirmatien of modeling adequacy could typically include comparisons of existing data obtained in large facilities selected contrasting of alternate calculational methods to portions of the code under consideration here, and establishment that scaling is adeoustely represented by the code. 1 3.2.7 Calculated Seabrock Thermal Response to Severe Accidents l Calculated behavie- to selected accident conditions has beer. sumarized by PSNW. Principal results and our coments are as follows:

1. Peak Steam Generator Temperature for loss of AC power and loss cf Feed Water Flow. The following temperatures and flow rates were calcu-lated at the indicated condition:

Location Temperature, OK Flow Rate, kg/sec Core (Peak) 1600 18 (recirculating between Upper Plenum 1160 upper plenum and core) Hot leg 760(wall) 2.4 (countercurrenti , SG Inlet Plenum 850 - l

SG Tube 700 (wall maximur) 3.3 (total in each direction) i SG Outlet Plenut 640 -

i 26

DRAFT FSF indicated tut tre hottest core rode wouid telt at about 30 seceret fror the time of these values, and that the generated hydrogen and blockage due to relocated core material would cause natural circulation betweer the core and the upper plenum to almost stop. At this point, the upper plenum would begin to cool due to energy transfer to the het leos. Plys (Ref.18) presents additional infortnation which shows temperatures continue to increase after vessel blowdown, with the peak upper plenum 0 temperature exceeding 1200 K for a short time. The tube temperature continues to increase for the time of the calculation (20,000 sec, with vessel rupture at 11,600 sec), reaching a maximum of about 10200K. We would be interested in seeing plots of other parameters over the span of the calculatior.s. including the hot leg and SG plena temperatures, to better understand the interactions and modeling. In response to a question, PSNH indicated they had not perforrred a detailed analysis cf reactor vessel hot leg nozzle thermal behavior, but felt a temperature of the order of 1000 K was necessary to cause failure. Discussion also identified that there was significant steam circulatory flow in the secondary side of the steam generator tubes, and that this steam, which was at a pressure corresponding to the stean generator safety valve settings, represented a significant hect sink. Further, it was an effective redium for transferring heat from hot tubes to colder tubes, thus tending to reduce the maximum tube temperature. This raises a question of what results would be obtained if the steam generators were depressurized to atmospheric pressure, thus maximizing pressure differen-tial across the tubes and simultaneously removing a heat sink which could influence temperatures throughout the NSSS. (A sensitivity analysis was conducted in which this was one of the parameters.) Infortnation presented in Reference 12 and the above sunenary table shows fluid flow rates in the het leg of roughly 2 kg/see as contrasted with a rate above 3 kg/sec in the SG tubes for the time after effective boiloff of water from the core until melt through of the reactor vessel. Cooling via stearr contained in the SG secondary side is thus an effective mediur for cooling the SG inlet plenum. The total mixing assumption pertinent U l l

DRArT tc fluic' in the plenu- it, in turr, effective in preverting het fluic' from reaching the tubes. This high tube flow rate is also effectivt in transferring heat from the reactor vessel to the SG secondary side, thus helping te lirit fluic temperature in the hot legs as well. We believe a st'udy would be beneficial of behavior with the SG secondary side depressurized after SG dry out. Now there would be no beat sink on the secondary side, and tube flow rates may be lower due to less of a driving force for natural convection flow in the SG. Further, we would expect to see further stratification in both the hot leg and the SG inlet plenum (the latter not being allowed in the PSNH supported analyses due to the noceling assurption of complete mixing). We pose the que: tion cf whether temperatures rm," bt significantly above what was calculated by PSNH and its contractors under these conditions.

2. Operater Induced Derressurization. This calculation was based on the assumption that the operator would open a PORV when the core exit themocouples indicated I?00 r. The calculations indicated accumulator discharge approximately 1400 sec after opening the PORV, with the RCS depressurized prior to vessel failure. The accumulators were emptied at about 10,600 sec, and vessel failure occurred 2000 sec later. Accumula-tor water was found to cause a small additional amount of hydrogen production. Phenomena associated with depressurizatien and hydroger decreased the effectiveness of heat transfer between the core and other regions of the NSSS. Steam generator inlet plenum temperature. reached a peak of roughly 8500 K du-ing the depressurization, then cooled, and U

remained below 650 K for the remainder of the calculation (20,000 see total calculation time, with PORY opening at approximately 8000 sec). 0 P.aximum tube temperature was about 650 K, and was reached at 20,000 sec, being identical to the inlet plenum temperature at that time. (Note RCS pressure is that of the containment following depressurization earlier in thecalculation.) We note that RCS pressure behavior (Ref. 18, Figure 4-4) is different for the base case and the PORV opening case prior to the time of opening of the POPV. We would like to discuss these differences for all partmeters 28 s

O b DPAFT and we w:Lic likt 10 urcerstand the reasor.s they exist. Ike nGte thert is little difference in temperature over the range in cuestion, and terperature is the importart parameter for the SGTR issue.) Volatile fission products represent about 20% of the decay heat, and the behavior of this energy source is calculated in the MAAP code. The , calculations illustrated movement of the decay heat source. About IP o' the decay heat was associated with fission products which were in the upper plenum at the time of vessel failure. A small amount was in the het legs, es was also the case for the pressurizer. The amount in the stear. generator tubes was not significant. (Most of the Csi was in the upper plenum at the time of vessel failure, with about 10% of the Csl in the hot legs.)

3. Other Variations and Uncertainty. Several sensitivity calculations W re performed to ottain a better understanding of b7havior. These included:

a. Higher core melt temperature

b. RCP seal failure

c. SG secondary side blowdown

d. Core resistance variation

e. Pecuced SG tube circulation

f. Core blockege changes.

These are discussed below.

a. Hioher Core melt temperature. A case was run in which core melt 0

temperature was assumed to be 3000 K as contrasted to the base case 0 2500 K. This was intended to delay the onset of core geometry degradation, which in turn provides more time to heat other portiert 0 of the RCS. The 500 K change in melt temperature was found to cause only a few degrees change in SG tube temperatures, which was attrib-uted to the extremely rapid temperature increase rate in the core as { melt temperature is approached, and a conconitant srall increase in the time available for heat transport to the steam generators.

                                    .                       29 l

l

. o DRAFT Tr.e r c'i' is based ur:r. assured symretric behavior, dereas so-( asyrxnetries have been found experimentally. If these contributed te a preferential flow of hot fluid near one of the hot legs, that leg right transport het fluid toward a steam generator and provide higher temperatures than determined in the calculation. This could increase the computed impact of the sensitivity calculation. A second aspect of the modeling that would act to reduce the calcu-latet impact of the sensitivity run is the assumption of mixing within the stean generator inlet plenum. We believe an assessmer.t of this effect is needed, as previously identified,

b. PCP seal failure. PCP seal failure, if it were to occur, was felt to be a leak in the range of 50 gpm (water) per seal. This was modeled, with the break occurring if, all four RCPs at 45 minutes after initiation of the accident. This was found to have an insig-nificant impact on the results (Pefs. 12 and 18).

PSNH also addressed preexisting leaks in SG tubes which are within technical specifications. These were stated to be small in compari-son to the 50 gpn flow rate associated with seal leaks, and conse-cuently were argued as being negligible (Ref. 17). We believe the preexisting leak situation has a negligible impact on NSSS behavior as long as the leak remains small, but do not accept the argument advanced by PSNH as the reason. A comparison of the velocity associated with flow in a tube due to natural circulation with that associated with the leak, with establishing that the latter was negligible, would be more convincing. Similarly, a comparison of flow rate induced by the RCP seal ruptore to that expected for natural convection flow would be helpful. Provision of temperature infomation, pertinent to fluid passing through the RCP seals would be helpful. l 30 l

DRAri

c. 55 seten:!r3 side blowccvn. Flys (Ref. 18) reports a celculation to irvestigate the effect of reduced cooling on the SG secondary side in which the steam generator PORVs are assumed to stick open, thus depletino the secondary side of a high pressure steam atro- '

sphere. Drastic differences were discovered early in the accident due to cooling as the steam generators blew down. Sufficient cooling was provided that the pressurizer emptied due to primary fluid contraction. Reactor vessel failure occurred slightly earlier in this case as contrasted to the base case due to less heat removal fror the primary system following removal of the secondary side heat sink. An initial peak in SG inlet plenum temperature of 8600 K is identical to that of the base case, but occurs about 500 sec earli-er. Following the initial peak, the plenun temperature behavior is similar to the base case, althouch displaced in time, but is 50 to 1000 K higher over the remainder of the transient. We suggest the calculation be conducted by assuming the PORV is stuck open after all water has been vaporized. This avoids the situation of overcooling associated with the early openino, and may be more corpatible with some postulated cperator actions associatec j with late attempts to deal with approaching core melt. k Again, we are concerned with the influence of assumed mixing in the steam generator inlet plenum and the impact upon calculated results.

d. Core resistance variation. Variation of the resistance of the core to flow was evatsated by lowering the axial and cross flow core' friction factors in one calculation. This slightly increased heat transfer to the steam generators and correspondingly increased tine to vessel failure. There was a slight tube temperature increase, but in general, the calculation showed little sensitivity of tube terperature to the change in core friction factors.

e. Reduced SG tube circulation. Selection of lower limit values of the number of steer generator tubes participating in flow from the inlet to the outlet plena was used for another sensitivity calculation.

31

DRAFT Tr.is proviced lower values of steam generater naturel circulatior flow relative to the het leg natural circulatier. flow rate, an:' reduced cooling of the steam generator inlet plenum due to flow from the cutlet plenum. Slichtly less heat was removed from the reactor vessel due to the lowered flow rates, and vessel failure occurred slightly earlier. These changes were insignificant. However, the steam generator inlet plenum was found to be about 1500 K higher ther for the base case, reaching a temperature of 9800 K for a short time. Steam generator tube temperature was relatively uraffected. Comparisen of inlet plenum and tube temperature transient bchavior (References 17 and 18's Figures 4-11 and 4-12) appears to indicate a sigrificant therral inertial associated with the. tubes, which do not increase in temperature to a significant degree in contrast to the temperature of the snurce fluid in the steam eenerator inlet plenum. ' We believe this needs further discussion. For example, what is the location of the tube temperature and does this location correspond to the highest tube terperature? Sezin, as previously stated, the influence of the assumption of complete mixing in the steam generator inlet plenum will impact the results. A portion of the concern is that reduced flow rates may lead to greater stratification and less mixing in the SG plena, a phenomenon that is not modeled in the PSNH reported evaluations, and a phenomenon with the potential to increase tube temperatures over what was rep 0rted.

f. Core blockace. In this calculation, a delay of blockage in the core at the time of core melt to the time the node was completely filled with refrozen eutectic was assumed. This was done to continue core oxidation and core / upper plenum flow for a longer time. For this case, the maximum sustained SG inlet plenum temperature is roughly 0

1060 K, with a short time (less than 50 seconds) temperature " spike" to about 11200K. 1 32 l

   . e DPtJT kt acair retterate th centerr. witt SG inlet plenur modeling and its impact upon the results.              '

g. Sensitivity Summary. An approximate comparisen of the results of the sensitivity study is provided in Figure 1. The major early effect on increased tube temperature is due to changing the SI: tube flow characteristics. later, and with the greatest impact, is the effect of delaying femation of blockage in the core, which allows continued circulation of hot fluid through the core where the temperature is increased, as opposed to a drastic reduction in heat transport between the core and other RCS components when a core geometry change occurs.

4 Steam Generator Tube Strenoth. Plys, in Reference 18, Appendix B, addresses SG tube integrity. The presentation is based upon the SG secondary side pressure being at the SG safety or relief valve setpoints which, as previously discussed, may not be the case. We note that Plys identifies nominal hoop stresses of 9300 to 10000 psi for the assumed conditions. Hence, the case of the SG secondary being depressurized will l result in a nominal hoop stress of roughly 19.000 psi. This stress, substituted into Reference 18's Figure B-6, results in a Larson Miller parameter of about 37. The Larson Miller parameter is defined as:  ; LM = T(20+1og t ) x 10'3 ' r Where* T = temperature, OP t = tire to rupture, hrs. r 1 0 Substituting a temperature of 1090 K (the value used by Plys to conclude the rupture time would be greater than P.5 hrs) yields a time to rupture of about 5 minutes, a significant change from the Plys value. l 0 Plys could have selected 1090 K as conservative, with no need to consider  ! en alternate since the no tube rupture position was supported by the 33

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. a DRAFT resf t. If n recogrize this possibility, and select a less conservative 10CN., we find a rupture tine of about 3.5 hours. These temperatures can be contrasted to the SG inlet plenum temperatures provided in Figure 1, with recognition that these are not tube temperatures, but also with the recognition that some of the parameters contributing to the temper-atures remain to be evaluated. Clearly, we are in a temperature region where relatively small changes have a significant impact upon creep rupture time. Equally clearly, tube stress could be roughly a factor of two higher than the value used to justify that tubes would not rupture. We conclude the picture is not as clear as presented in Reference 18, which presented a conclusion that tubes would not be ruMured. 3.2.8. Other Considerations In Reference 17, PSNH stated that if one postulated creep rupture failure of steam generator tubes, the pressure inside the previously dried out and isolated steam generator secondary side would incmase until the steam generator PORVs setpoint was reached, at which time the valves would lift and modulate until reactor vessel melt through and RCS depres-surization into the containment. During the periods of SG PORV opening, there would be a high leak rate bypass condition directly from the RCS to outside the containment. They further stated that after vessel melt through, the leak rate out this path would be low and would correspond to any low pressure leakage through the reclosed PORV. They note this leal path could be enhanced if the SG safety valves also lift and fail to reseat properly; however, they believe it unlikely that the safety valve setpoint would be reached. As previously discussed, we do not believe an individual tube would rupture, but instead believe there would be a massive failure in one steam generator. (Once the failure initiated, we would expect the RCS to depressurize rapidly, which would reduce stress on tubes in other steem generators. ) It is difficult to postulate a p0RV modulating this condi-ticr. It is further difficult to postulate the PORV or the safety valves 34

, a DRAFT woulc net be darage:' wher exposed to these conditions, and therefort their reclosing may be cuestionable. Finally, if the conditions which led to the accident secuence involve a loss of all AC power, which is one of the likely situations oiven a severe accident scenario, we pose the question of how long the PORVs can be expected to modulate pressure assuming they are not damaged by the fluid being modulated. Plys (Ref. 18) has identified that the MAAP code does not model certain aspects of SG tube ten;perature, and a method of obtaining temperature was discussed. Aside from the impact of secondary side stear as a cooling mediu'r, we are concerned about local heating due to small leaks. Such a leak could cause a small amount of hot fluid to pass through a localized area into the SG secondary side, with different heat transfer character-1stics and tube temperatures than one would encounter with the treatment of overall inside to outside heat flow utilized by Plys in their estima-tior. Whether this is important to localized tube temperature over a sufficiert area to be of concern should be addressed. (Note the effect could also be concentrated in an adjoining tube. This can be visualized by picturing a tube with a small hole which directs het RCS fluid onto the secondary side surface of an adjoining tube, while the inside surface of that same tube is exposed to het RCS fluid.) 3.3 Accident Likelihood PSNH has estir.ated the mean annual frequency of accidents in which the core nelts with the RCS at high pressure and the SGs dry as bounded by a value of 4.5 X 10-5 per reactor year (Ref.17). This is composed of the following plant darage states: 35

. w DRAFT Flatt Cari;. Mear: Annual State (PDS) Frequency 3D 1.5 X 10-5 3Fr 8.9 X 10-6 4A 1.4 X 10-5 4C 1.7 X 10~7 4D 2.8 X 10 II 4E 2.2 X 10'7 4FP 1.2 X 10~ 8A 3.9 X 10-6 Total 4.5X 10 -5 The accident sequences which comprise the PDSs include transient and loss of off site power sequences with failure of all emergency feedwater, failure of feed and bleed with loss of all emergency feedwater, and transients without scram. PDS 8A consists of eight secuences which involve station blackout and emergency feed water failure with recovery of containment heat removal. PSNH also addresses the potential impact of tube rupture on this information. They have assigned a high chance of no containment failure to PDSs A. PDSs C and D are considered as leading to a high likelihood of long term containment overpressure failure. PDSs FF are a high chance of small bypass, and PDS E is a high chance of large bypass. Hence, PDSs A, C, and D would be impacted by SGTR, and FP may represent some impact. Addition of the appropriate valuet indicetes that the likelihood of being in a condition where SGTD could affect the results is about 4 X 10-5 (as contrasted to the assumption of no SGTR). PShH considers these values to be bounding because some of the values incluce states with water on the steam generator secondary side, for which SGTR is not a concern, certain operator recovery actions have been neglected, and pCS deprt.ssurizations prior to core melt have not been considered. As previously discussed, operator depressurization 1; one of the potential steps which one could consider to mitigate SGTR. PSNH estimates the frequence of operator failure to depressurize as less than 10-2 to 10-3 per demand, provided proce-dures are modified and adequate operator training is provided. These values lead to a cenclusion that the frequency of obtaining conditions under which SGTR would be of concern can be reduced to of the order of 10~7 to 10-P per reactor year. 36

 . s DRAFT Althcugh these values appear reasonable, we note that the conditions which led 2

to the factor of IV to 10-3 reduction do not presently exist. We further would need substantiation for these values prior to acceptance. Discussion is also provided concerning the likelihood of SGTR if exposed to high pressure core melt conditions (Ref. 17). PSNP points out that their calculations shew SG tube temperatures that are roughly 200 to 300 0F below what would be required for creep rupture, and this is identified as principal-ly due to conlang by steam on the SG secondary side. Several things are necessary for teceptance of the tube temperature conclusions, including, as discussed elsewhere, substantiation of the calculational technique and inves-tigation of the likelihood of the SG secondary side having a significant steam inventory (which also means having a significant pressure). Finally, PS M estimates a 991 chance that failure of SG tubes will not occur before reactor vessel melt through or piping nozzle failure. This value, combined with the prior estimates o' frequencies, appears sufficient to establish that SGTR is not of concern as a significant contributor to risk. Therefore, one can reasonably anticipate that substantiation of the various iters which led to the conclusier., as discussed in this comunication, should provide substantiation of the above preliminary conclusion. 3.4 Additional Peviewer Observations A number of observations and connents have been made in the previous discus-sien. We offe" the following additional coments:

1. Much of the modeling utilized in the calculations has not been document-ed. We understand this is underway. Such documentation will be helpful I in the continuation of the review.

2. The outside of the het legs is assuned to be adiabatic. This probably introduces a srall conservatism into the results with respect to hot leg temperature. The impact on other parameters is probably negligible.

With respect to the hot legs, the parameter of interest may involve a relatively thin wall connecting pipe that is exposed to high fluid 37 i _

 , o DRAFT ter;erature, and w' sc terperature will follow fluid temperature more closely ther. is the case with the relatively nassive hot leg; or the vessel nozzle region of the hot leg, which will be more closely allied with fluid circulating rapidly within the upper plenur. Thermal response of these regions may be critical in determination of the failure point of the RCS pressure boundary,

3. Although the limited experimental evidence reveals some symmetry in flow behavior within the reactor vessel, there are also unsymmetrical flows and temperatures. We understand the MAAP calculations are based upon nedeling the upper plenum fluid as a single volume. This appears to be a nonconservative approach.

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                                                ?P 1

o - DRAFT 4 1 STEA" GENER T TUEE FUrTUTE CONCLUSIONS j The above discussed considerations lead us to the conclusion that this topic is in a develeping state, with knowledae being rapidly accurulated. Insuffi-cient infonration is presently available for one to conclude that SGTR cannot occur as a result of severe accident conditions. Our judgement, at this juncture, is that a carefully conducted and thorough evaluation on the part of PSNH, that utilizes infonnation which either exists or will be available within the near future, can establish that the likelihood is strall that a SGTR will result due to overheating during severe accidents. Further, our judgemert is that the risk associated with SGTR can be shown to be negligible fer these cc.ditions. Our judgement needs to be substanti-ated. We have encountered too many unanswered questions, unsubstantiated assumptions, and potential conditions which could lead to calculation of increased temperature to accept a conclusion that SGTR will not occur under circumstances such that the associated risk can be neglected. We note, as a cualifier to these conclusions, that our review is not complete, and, in addition, work is ongoing to provide further infonnation. Existing knowledge would support a conclusion that SGTR is not a problem if the RCS is depressurized. Consecuently, reasonable assurance that progressions toward core nelt would not occur at high RCS pressure, coupled with suitable technical backup for a conclusion that low pressure is not of concern, would eliminate our concern regarding SGTR under severe accident conditions. We have not conducted an evaluatior. of the trade-offs associated with such an approach, nor have we been provided with infonnation that would either support ' or negate RCS depressurization under severe accident conditions. We have not I provided 6 recorrendatier regarcing whether RCS depressurization is attractive when all pertinent factors are considered. 39 l l

e DRAFT j Substar.tietier c' e judeerer.: that SGTF is not a concern unde

• severe acticer.:

l conditions with the RCS at high pressure can be obtained through a combination of analytic and experimental investigations. The ongoing test at Westinghouse in which reasonably close similitude is claimed between the test facility and appropriate parts of a Westinghouse four loop NSSS should provide key data which can be applied to assist in the confirmation of analysis technioues. Selected test data from other facilities and further examination of the analysis techniques, coupled with necessary changes when they are uncovered, should provide sufficient confirmation that reasonable reliance can be placed upon accident analyses pertinent to this issue. Application of a reliable analysis technique to issue investigation should then provide the necessary background to resolve this issue. 40

DRAFT E. PEFEDE*TES

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2. Garrick, John B., Karl N. Fleming, and Alfred Torri, "Seabrook Station Probabilistic Safety Assessment, Technical Summary Report," Pickard, Lowe and Garrick, Inc., PLG-0365, June 1984.

3. *Seabrook Station Risk Management and Emergency Planning Study", Pickard, Lowe and Garrick, Inc., PLG-0432. December 1985.

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7. Denny, V. E., and B. R. Sehgal, " Analytical Prediction of Core Heatup/Licuefication/Slumpirg," Paper TS-5.4, Proceedings Intl. Meeting on LWR Severe Accident Evaluation, Cambridge, MA, August 28 - September 1, 1953.

8. Lyon, Warren C., " Report on Meeting to Discuss RCS Pressure Boundary Heating During Severe Accidents (May 14, 1964)," NRC Memorandum to Distribution, June 15,19E4

9. Eernero, Robert M., "Need for Multidimensional Modeling of RCS Behavior in Support of Severe Accident Investigation," NRC Memorandum for Denwood F. Ross, August 30, 1984

10. Sheron, Brian W., " Steam Generator Tube Response During Severe Acci-dents," NRC Memorandum to B. D. Liaw, February 14, 1985.

11. Stewart, W. A., A. T. Piecrynski, and V. Srinivas, " Experiments on Natural Circulation Flow in a Scale Model PWR Reactor System during Postulated Degraded Core Accidents," Paper 10.C. Proceedings of Third International Topical Meeting on Reactor Themal Hydraulics, Newport, RI, October 15 - 18, 1985.

12. Plys, Martin G., Marc A Kenton, Robert E. Henry, and Peter Kirby, "Seabrook Steam Generator Integrity Analysis," Information presented at Brookhaven National Laboratory by Fauske & Associates, Inc. and Westing-house Electric Corporation, October 17, 1986,.

13. Stewart, W. A., A. T. Piecrynski, and V. Srinivas, " Experiments on Natural Circulation Flow in a Scale Model PWR Reactor System During Postulated Degraded Core Accidents," Westinghouse Electric Corporation, Pittsburgh, PA, Scientific Paper 85-5J0-PCIPC-P2, August 29, 1985. i 41

I DRAri 14 Stewart W. A., A. T. Pieczynski, and V. S-inivas, "Experinents on Natural Circulation Flows in Steam Generators During Severe Accidents." Westinghouse Electric Corporation Pittsburgh, PA, Scientific Paper 85-5J0-RCIPC-P3, Decer:ber 5,1985. 15. Chen, E. C-J., H. M. Doranus, W. T. Sha, and R. R. Sehgal, " Degraded Core Study using the Multidirnensional COMMIX Code," Trans. ANS. Vol. 49, pp. 453-454, June 1985.

16.

Dearing,

J. F., " Flow-Pattern Results for a TFLB' Accident Sequence in the Surry Plant Using MELPROG," Los Alamos National Laboratory, LA-UR-85-3668, November 1985.

17. 'DeVincentis, John, " Response to Request for Additional Information (RAIs) " Letter from Public Service of New Hampshire to Steven M. Long of NRC, SBN-1227. T.F. B7.1.2, November 7,1986.

16. Plys, Martin G., et, al., "Seabrook Steam Generator Integrity Analysis,"

Fauske & Associates, snd Westinghouse Electric Corporation, November, 1986. (Provided via Reference 17.) I 4 4? l}}