ML20065D824
| ML20065D824 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 09/17/1982 |
| From: | Budnitz R, Davis P, Fabic S FUTURE RESOURCES ASSOCIATES, INC. |
| To: | |
| Shared Package | |
| ML20065D822 | List: |
| References | |
| NUDOCS 8209300262 | |
| Download: ML20065D824 (229) | |
Text
{{#Wiki_filter:- - . - _ _ _ . REVIEW AND CRITIQUE OF PREVIOUS PROBABILISTIC ACCIDENT ASSESSMENTS FOR TIIE SIlOREHAM NUCLEAR POWER STATION Report on Task 1 of the Project
" Consequence Assessment for Suffolk County Radiological Emergency Response Plan" VOLUME I: MAIN REPORT Prepared for Radiological Emergency Response Plan Steering Committee Frank Jones, Chairman Office of the County Executive County of Suffolk II. Lee Dennison Building llauppauge, New York 11788 Robert J. Budnitz Principal Author Peter R. Davis Stan Fabic Howard E. Lambert Contributing Authors September 17,1982
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o REVIEW AND CRITIQUE OF sq ! - 4 a PREVIOUS PROBABILISTIC ACCIDENT ASSESSMENTS FOR THE SHOREHAM NUCLEAR POWER STATION _- i Report on Task 1 of the Project _ " Consequence Assessment for Suffolk County Radiological Emergency Response Plan" j VOLUME I: MAIN REPORT - Prepared for 2 Radiological Emergency Response Plan " Steering Committee o Frank Jones, Chairman Office of the County Executive County of Suffolk H. Lee Dennison Building Hauppauge, New York 11788 - Robert J. Budnitz Principal Author ] Peter R. Davis Stan Fabic _ Howard E. Lambert j Contributing Authors a 4
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September 17,1982
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TABLE OF CONTENTS l l 1.0 Introduction 1.1 Overall Conclusion 1.2 Limitations 1.3 Objective of the Report 1.4 Organization of the Report 1.5 Summary of SAI's Results 2.0 Are All Important Accident Sequences Considered? 3.0 Are the 5i.I Calculations of the Accident Probabilities Correct? 3.1 Methodology Considerations 3.2 Data Base Considerations 3.3 Specific Accident Sequences -- An Eclectic Review 3.4 Implications of the Review of Specific Accident Sequences 4.0 Are the Accident Phenomena Within the Plant Treated Correctly? 4.1 Grouping of Accidents Into Classes 4.2 Review of Phenomena After Meltdown 4.3 Some Specific Issues Involving Meltdown and Related Phenomena 4.4 Core Vulnerability vs. Core Melt: Containment Event Trees 4.5 Implications of the Review of In-Plant Phenomena
D l TABLE OF CONTENTS ! (continued) ) 5.0 Summary and
Conclusions:
Are the Probabilities and Magnitudes of the Calculated Potential .!adiation Releases Correct? ) 6.0 Acknowledgments 7.0 References J Appendices A. " Scope of Work," excerpts from the Contract between Suffolk - County and Future Resources Associates, Inc. B. " Review of Within-Containment Phenomena and Containment Challenges," report by Dynatrek, Inc. under subcontract to this - project J C. "A Selected Review of the PRA Prepared for the Shoreham Nuclear Power Station," report by P. R. Davis prepared as part of this project D. " Internal Flooding Analysis," prepared by H. E. Lambert as part of this project D D D D D D
'O 'O 1.0 Introduction O This report sumarizes the work carried out by Future Resources Associates, Inc. (FRA) under contract to Suffolk County, New York on the project' entitled " Consequence Assessment for Suffolk County Radiological Emergency
- O Response Plan." The overall goal of the project is to provide Suffolk County with technical support in its development of an emergency response plan for the Shoreham Nuclear Power Station, in particular by providing O technical input as to the probabilities, severities, and radiological dispersion characteristics of potential large accidents at Shoreham.
Shoreham is a boiling water reactor (BWR 4-Mark II) in final stages of O construction on the north shore of Long Island, New York, with an electrical gross power rating of 846 megawatts. The reactor was manufactured by the General Electric Company and the architect-engineering C work has been done by Stone & Webster. The location of the reactor on Long Island and the site plan for the Shoreham facility.itself are shown on the next two pages: these are reproduced directly from Figures 2-1 and 2-3 of O the NRC's " Safety Evaluation Report" for Shoreham (Ref. 6). This project is a joint one involving a single unified scope of work under C two contracts, one' with FRA and the other one with Finlayson & Associates of Cerritos, California; under the arrangement with Suffolk County, Fred C. Finlayson of Finlayson & Associates has assumed overall responsibility to O O
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4 O coordinate the two contract efforts. The unified scope of work of the two
- O contracts is reproduced in Appendix A of this report, and covers four
! tasks. FRA, under the technical direction of Robert J. Budnitz, has been principally responsible for the work under Task 1, while Finlayson & O Associates is principally responsible for Tasks 2 and 3; in each task the other party has played a supporting role. Task 4, involving project management, integration, documentation, and technology transfer, has been a O shared responsibility. The contract was signed in late June 1982, and work under it has taken O place predominantly in the months of June, July, and August, 1982 with this report due in draft form on September 15. O The review team made a site visit to the Shoreham plant in late June. Preliminary discussions of findings have taken place on an almost continuous basis, informally and verbally, between Dr. Budnitz and Dr. O Finlayson in order to assure that both parts of the overall project are integrated effectively. Both parties agree that the integration has beer successfully accomplished. O The FRA responsibility within the study has been predominantly to carry out Task 1, " Review and Critique of Previous Probabilistic Risk Analysis," and C this report covers the work that FRA has accomplished in carrying out Task
- 1. Task I has consisted largely of a review of the preliminary draf t report coordinate the two contract efforts. The unified scope of work of the two O
O
C) 5 entitled "Probabilistic Risk Assessment, Shoreham Nuclear Power Station," O that was carried out by the San Jose, California office of Science Applications, Inc. (SAI) for Long Island Lighting Company (LILC0), the owner of the Shoreham facility. To carry out this review task, FRA has O utilized a team of four individuals who possess expertise in various aspects of water reactor safety and in particular of probabilistic risk l ssessment-(PRA). The FRA work has been led by Robert J. Budnitz, President O of FRA, and has included Howard E. Lambert and Peter R. Davis, FRA consultants, and Stan Fabic of Dynatrek, Inc. (Rockville, MD), an FRA subcontractor. O The scope of the FRA work in Task 1 has been that of an independent review, which must be understood as quite different from an independent analysis of the potential accidents at the Shoreham plant. The purpose of the review has been to ascertain whether the PRA results obtained by SAI in their large and voluminous s+udy are sufficiently reliable to form an acceptable basis fcr the County's emergency planning work. Because the scope of work has not included significant independent analysis, it is important to realize that its conclusion cannot be considered as a
" stand-alone" conclusion. That is, it depends heavily upon the quality of the detailed work by SAI. This point was made in the original proposal to the County, where it was pointed out that "if major flaws are indentified n"
in the earlier studies, . . . it may be necessary to devote a higher level of effort to this project." Thus the basis for the project has been an assumption that the SAI work under review is a credible effort, up to the O standards of the state-of-the-art, and requiring no major upgrading. The O
O 6 FRA team has attempted to challenge this assumption by carrying out a O critical review with the intent to uncover facets of the SAI analysis that might contain inadequate methodology, inappropriate assumptions, errors of omission / commission, or biases. O 1.1 Overall Conclusion The assumption that the SAI work is basically sound has turned out to be correct, in the opinion of Future Resources Associates, and we believe that the following conclusion is essentially the most important overall sumary of the results of our study: FRA concludes that the overall results of the Shoreham PRA, contained in the preliminary draft report by Science Applications, represent reasonable conclusions as to the likelihood and O magnitude of releases from large accidents at the Shoreham reactor. As discussed in the text below, we believe that the likelihood of core melt O accidents is somewhat higher, and the magnitude of radiation releases somewhat smaller than found by SAI. However, we believe that the differences should not cause Suffolk County to modify their emergency O preparedness activities significantly compared to what they would do by using the SAI results directly as published. Subsidiary to this important conclusion are our conclusions that the methodology used in the O SAI study is at the level of the state-of-the-art of reactor risk assessment at the present time, and that SAI's application of this methodology to Shoreham has generally been a competent onel. O O
m 7 a 1.2 Limitations 7 There are limitations to our acceptance of the SAI study, and these are based on limitations within the study itself. The most important of these 3 study limitations, from the point of view of its applicability to the County's emergency planning needs, is the absence of any analysis of. accidents arising from the set of " external initiating events" that could 3 initiate accidents from outside the plant. Most important of these external initiators is earthquakes, with high winds (hurricanes, tornadoes) and floods also of possible concern. In addition, internally-initiated 3 fires are not treated. Because these types of accidents could be important risk contributors, their omission means that neither LILC0 nor the County has a fully satisf actory set of important accidents to use as a basis for
- 2) emergency planning.
There are other limitations to the SAI study, including the appro imations, I) conservative in nature, as to the fission product source term; the way the very numerous accidents were grouped into classes for ease of treatment in the within-plant and containment phenomenological analysis; the absence of II a plant-specific f ailure data base (since Shoreham has not yet operated); and other issues involving methodological approximations made by the study team. Fortunately, it is our view that none of these limitations, with the II exception of the conservative approximations as to the fission product source term, is sufficiently important that it might significantly alter the substance of the County's emergency planning effort. In any event, 9 O
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. - _________________._____________.____._..-__.s
8 O while we found a few places where the SAI report could be improved, we also O believe that many of the report's limitations are not SAI's limitations per g but rather are limitations of the present state-of-the-art of PRA generally. O Our own study within Future Resources Associates, unfortunately, has had limitations of its own. The most important of these has been the relatively O short time period (about 2-1/2 months) and relatively small level of effort that has been devoted to the project. Although the contrict has been sizable from Suffolk County's perspective, the total level of professional O effort devoted to reviewing SAI's draft PRA report has been only a fer percent of the effort that SAI spent on carrying out the PRA itself and, of course, such a few percent effort cannot reasonably be expected to study O every f acet of the problem. This limitation is, however, balanced in part by the fact that the review team has had considerable experience in the PRA field, which has enabled the effort to be focused on what are thought 'O to be the main issues. Another limitation that hampered our group's work during the first month, O but was cleared up in mid-July, was a restriction on us that effectively barred direct contact with the SAI PRA study team. After it was lifted, our interaction with the SAI team was a full and open technical exchange of O questions and issues, at the level of professional mutual respect that we found refreshingly matter-of-fact. We wish to acknowledge SAI's full cooperation, for which we are grateful. O O
O 9 There are substantial uncertainties associated with the numerical O conclusions that SAI quotes in its report. The uncertainties arise from several sources, including the validity of the data base, approximations in the accident sequence fault-tree / event-tree modeling, uncertainties and O gaps in our understanding of within-plant accident phenomena, and incomplete understanding of the role of human error and human ingenuity in reactor accidents. Our general view is that the treatment of uncertainty O within SAI's study is a reasonable one, including as it does advanced methods for estimating contributions from various sources. We have concentrated on those uncertainties that could particularly affect what O Suffolk County might do or decide in the context of its emergency response plan. The discussion in the main body of this report will be in that context. O 1.3 Objective of the Report 'O It is important to state clearly the objective of FRA's review work, which has been to provide Suffolk County with an independent technical \O opinion as to the probabilities and magnitudes of large potential accidents at Shoreham. FRA's work under Task 1 of its contract has concentrated on ascertaining, through independent technical review, whether the preliminary O draf t version of the LILCO- supported probabilistic risk assessment carried 1 out by SAI provides a technically sound basis for emergency response planning. 'O O
i l D 10 1.4 Organizatica of the Report
'3 The body of the report to follow will be organized generally along the lines of the scope of work (see Appendix A) for Task 1 of the overall C) project.
The technical issues that we have covered within the SAI report can be C) conveniently separated into the following:
- Are all important potential accident sequences considered?
C) - Are the calculations of the accident probabilities correct?
- Are the accident phenomena within the plant treated correctly?
- Are the magnitudes of the calculated potential radiation releases C) correct?
C) The treatment of environmental transport of radioactive materials af ter release, and their impact on populations, is the subject of Tasks 2 and 3 of this study and is not covered in this report. O 1.5 Summary of SAI's Results C) Before continuing with the results of FRA's review, it is useful to reproduce here the main results of the SAI study, in tabular and figure form. These, on the following pages, are reproduced directly from the SAI C) - 0
O- 11 l draft report. The first page reproduces SAI's Table 4.1 and Figure 4.1,.in l O which the frequencies of core vulnerable conditions are shown. Of particular interest are the uncertainty ranges of the SAI results. The second page reproduces SAI's Table 4.2, which contains the detailed O numerical results for SAI's five accident classes. Core vulnerable frequencies are shown along with contingent probability of core ;.elt, and the characteristics of the releases are also shown. The details of Table
- O 4.2 will be discussed below in the body of our review.
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12 O (reproduced directly from sal's report on Shoreham) l DRA7.pggy~ ~"ngy l Table 4.1 I
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14 O 2.0 Are All Important Accident Sequences Considered? O The answer to this question, as mentioned above in the introduction to this report, is negative. In particular, there has been no treatment of
- O internally-initiated fires, nor of any externally-initiated events, the most important of which are earthquakes, floods, and high winds (hurricanes, tornadoes). These omissions are discussed in the SAI report l
lO (P.1-13), and were beyond the scope of that study. The implications of these omissions will be discussed below, but first we will consider whether any important internally-initiated accidents (besides fires) seem to have C been omitted. On this latter point, we have concluded that the SAI draft report has O apparently considered all of what the reactor safety comunity considers the important internally-initiated accidents. Specifically, we have not found any internally-initiated sequences likely to contribute significantly O to the overall risks that have been omitted. Of course, the unconscious omission of important sequences known to others O in the reactor safety community would be quite unlikely in a study of this kind . . . the safety community maintains close enough and open enough communications that any new or unusual accident sequences would almost O surely have cometo the attention of the study team or its outside consultants. So our conclusion comes as no surprise. O This observation of the apparent co.npleteness within sal's draf t report does not mean that there are no,important sequences omitted. It only O
O 15 means that we are unaware of any, nor do we believe that anyone else in the O reactor safety community has thought of any. The strength of our convictions about completeness is based on the general observations that, over nearly a decade of time since the initial probabilistic accident n" delineation work of WASH-1400 (Ref.1), there has been hardly any ad'lition to the list of important internally-initiated accidents that WASH-1400 considered. But, on the other hand, who knows whether or when a new
*a sequence will arise, either from operating experience or from analysis?
The grouping into sequences that the SAI team used is discussed in detail O in Section 4.1 later in this report. The five classes are described in SAI's Table 3.3.1, which is reproduced on the next page.
.O It is important to recognize in this connection, however, that the approximately 1600 reactor years of commercial nuclear power operating experience worldwide without a serious accident leading to off-site O consequences is a statistically significant data base providing evidence that the result for Shoreham is unlikely to underestimate the probability of serious accidents by a large factor, unless Shoreham is somehow very untypical in its risk profile of the entire group of reactors; of courses it is just this question that is being addressed by the Shoreham-specific risk assessment.
O Returning to the known " external" omissions (fires, earthquakes, high winds, floods), the SAI team acknowledged in their report (P.1-13) that
;O O
16 lO O
~O (reproduced directly from the Shoreham PRA by SAI)
Table 3.3.1 GENERIC ACCIDENT SEQUENCE Ct. ASSES lO l ' atPRtitMTATIVE 3Y37D8 Ltytt PMT5! CAL SA115 St0VENCE FDA CLAS$ GEnte!C ACC!Ctnf C!Bitt!BUTING EVENT SEQUthCE FOR CLA511FICAT10N SEQtt%CE DL11GMATOR Transtants involving less of Transient with less of Relatively fast care setti in,e,te,y meteo,, e,eli. ni,n end,io, ,, esso,e . ClassI(C1) contain,ent intact as core 'O melt and at few pressure emell LOCA events involving toelant makews less of inventory askeve; transtants involving less of scree function sad tesellity te prestee sufftstent caelant I askewe Traniteats or LOCAs in 1ving Transient wita insi ef asiatively slew core melt resideal heat remval ' O Ciani it (C2) dee ts fewer decay neat less of containment heat re. naval inadvertent SRV epecing op ers containment failed accidents vita inadequate nest prior to care melt removal capestitty l Large LOCAs witn insuffIctent Large LOCA witn less nelatively fast core molt; of low pressure ECC5 Cle,ss 111 (C3) entatamat intact at core coolant makewir. transients utth aalt but at high internal less of heat removal and long. ,g teru less of inventory askeupg pressure-i RPV fattures vita insufficient toelant askeup Transtants involving less of Treasient with fatture Relatively fast core molt of RPS and fatture of Class IV (Cs) containment fetis prior te scree fvaction and less of ef SLCS care melt due te everpressure containment heat removat er all reactivity centrols transients witn tess of scre. fonction . O fellowed by actuated espressuri. 34 tion LOCAs outside containment with LOCA (n main steen Relatively fast core melt; lines with fatlere Class V (CS) contairment failed free ini. insufficient coolant askews te of sq51V closure and core;RPV fattures meten result atten of accident due to in saaestate centstement f atture less of ECCS equipment fattures O O f O
O 17 these were consciously and specifically excluded. The reasons for the O omission are probably a combination of two things: first is the fact that treating external events and fires in a PRA is considerably different than treating internal initiators, requiring a different nethodology, large O additional manpower resources different in character from the rest of the study, and yielding results of even greater uncertainty than the uncertainty in the rest of the PRA analysis. These large uncertainties are O because the methodologies ar'e immature and the data base weak. Second, the SAI study team believed when the project began in 1980 that these external events and fires were not as important contributors to risk as D internally-initiated accidents, nor as amenable to cost-beneficial risk reduction (Ref. 2). In the intervening two years, our ability to analyze the risk contribution from fires and externally-initiated events has advanced considerably. Benchmarks of this advance include the recent publication by Pickard, Lowe,
& Garrick of seismic and fire PRAs at both the Zion and Irdian Polat reactor stations (each a two-unit station with Westinghouse PWR's); the completion of an important NRC-sponsored seismic methodology development
~ O effort, the Seismic Safety Margin Research Program at Lawrence Livermore National Laboratory; and the inclusion by consensus within the draft PRA Procedures Guide (Ref. 3) of an " acceptable" methodology for PRA analysis 'O of earthquake- and fire-initiated accidents. With these methodological advances have come the first probability-based insights into the quantitative contribution to risk from these sources, albeit with very
- O O
O 18 large uncertainties and important conservatisms in the analyses. To the O surprise of some in the reactor community, the Zion and Indian Point studies have told us that neither earthquakes nor fires can be neglected as contributors at those power. stations to overall public risk. Their
'O apparent preeminence at Zion and Indian Point is partially because internally-initiated accidents were found to be much less important than was found in WASH-1400.
O Fortunately for the purposes of this report, one member of the PRA review team (Budnitz) has recently been reviewing the Indian Point PRA O specifically rrom the perspective of the risk posed by earthquakes and high winds (but not fires). His basic conclusion vis-a-vis earthquake-initiated and wind-initiated accidents at Indian Point is that the O methodology is clearly adequata to tell us what types of accident sequences are likely to be of most concern, and whether these initiators pose important safety problems (at Indian Point, they do). But th'e O methodologies da not seem to be mature enough to provide reliable quantitative calculations of the probability of core melt . . . thus any numerical comparison of internally-initiated core melt frequencies or j O public risks with those from earthquakes and winds is of little value. ( An important observation from the Indian Point study is that accidents f0 initiated by earthquakes and winds seem to involve phenomena quite similar to those involved in the ensemble of internally-initiated accidents: that is, the accident phenomena themselves do not seem to form a different set iO l lO l L
O 19. of phenomena that must be considered separately in the sense of comprising n If this observation is generally valid, then
" different types of releases.
the main impact on overall PRA results would be to increase the probability of releases already treated in the analysis, with less impact O on the spectrum of releases and consequences. What insights about Shoreham can be obtained by transfer to Shoreham of our O recent increased understanding of externally-initiated accidents at other plants? Unfortunately, not much. This is in part because there is not yet available any external-events PRA analysis for any BWR reactor (the PWR's analyzed to date are in detail not remotely similar to the Shoreham design), and in part because the accident-initiating events found to be important at Zion and Indian Point are quite site- and design-specific in detail, involving features that are unlikely to be reflected at another plant. ,0 Much thought has been given by the PRA review team as to why useful ! insights applicable to Shoreham cannot be reliably gained from studying external-events PRA analyses at other reactors. Our negative conclusion arises basically from our belief that if earthquakes, high winds, or fires give rise to important accident sequences at Shoreham, the sequences themselves are likely to be idiosyncratic to Shorehtm, or in some cases n" possibly generic to BWR's (or BWR Mark 11 reactors) as a class. Absent any specific analysis, we conclude that the contribution of these sources to residual public risk from Shoreham is simply not quantitatively known, in ( O terms of either probability or character of consequences. l O l
) 20 The insight (at the PWR's studied elsewhere) that the accident phenomena 3 are not qualitatively different in kind from those arising from internal sources is, of course, a reasonable one consistent with the intuition of most students of the problem. If this insight were to hold at Shoreham, 3- then emergency response plans based on accident scenarios from internal initiators would likely afford reasonable protection from these other types of accidents as well, provided that the special external circumstances ) surrounding a large earthquake or hurricane are adequately included in any response plans, but the "if" in this proposition could be a weak reed. J D D D J J 3
O 21 3.0 Are the SAI Calculations of the Accident Probabilities Correct?
!O To answer this question, the FRA team reviewed the methodology used by the SAI analysts, considered the validity of the numerical data base, and g ~
repeated selected calculations to ascertain how sensitive some of the results were. Because the numerical quantification of literally hundreds of sequences, within many different event trees, was beyond our capability, O we are not able to affirm the specific validity of each SAI accident sequence. However, we believe that such a detailed review has not been necessary to satisfy Suffolk County's objective.
- O 3.1 Methodology Considerations O
The methodology of the SAI analysis includes numerous advances over the WASH-1400 analysis of Peach Bottom in 1973-1974. We concur in the 'O judgments of the SAI team that the use of these advances improves the an alysi s . An example of the different approach taken at Shoreham is the incorporation of certain important support systems (such as instrument air,
- O AC power, DC battery power) consistently within the fault trees rather than having some of these in the event trees, as in WASH- 1400. This approach allows for easier analysis, but carries with it an increased risk that
.O common dependencies might be missed through oversight if the analyst gets sloppy or forgetful. We believe that SAI's approach is a valid one, which can produce valid results if executed with care. !O l lO
-O 22 Another advance is the differentiation in the Shoreham study between D accidents leading to a " core-vulnerable" condition and those that proceed further to " core melting." In the WASH- 1400 approach, any accident proceeding as far as what SAI here calls a " core-vulnerable" condition was assumed to proceed all the way to " core melt"; that is, there was no differentiation in WASH-1400 between the two conditions and the more conservative approach was taken. The reason for this was that in 1973-74 0 there was no existing methodology for consistently calculating the of fferentiation between these two conditions, and the WASH-1400 study team was not able to develop one within the resources and technical knowledge of G' the time. Now, eight years later, this differentiation is feasible, and the SAI analysts have developed it further and applied it in the Shoreham analysis. O In our opinion, this differentiation is an advance: it was understood in the WASH-1400 period that the conservative core melt assumption was not O' correct, and the Three Mile Island accident told us this as well. (The TMI accident, if treated with the WASH-1400 methodology, is called a full " core melt," although it was certainly not.) For purposes of comparison, the O probability of " core-vulnerable" conditions at Shoreham is what should be compared to " core melt" in WASH-1400. The numerical results of the Shoreham analysis show that the fraction of " core-vulnerable" sequences that proceed g to " core melt" ra..ges from several percent (for what are called Class I and II events) to 100% (for the Class V category). In Section 4.4 below, we will discuss in detail the question of whether SAI's quantitative 'O conclusions in this area are valid. O
O 23 Still another methodological advance is in the way uncertainties are O estimated. Since the WASH-1400 team pioneered with their analysis, a variety of different methods have arisen for getting a numerical handle on these uncertainties. The approach taken by SAI in the Shoreham study is to
.O use Monte Carlo methods to model how much difference in the final results would arise from various changes at the front end, or in the data base.
This seems to be a reasonable approach, and its conclusions are also quite O reasonable (see Table 3.8.1 on Page 3-172 of the SAI main report), but we did not review it in detai. because of our judgment that the conclusions were reasonable. O Of course, there are still methodological problems that temper the confidence we might have as to the validity of PRA results for frequency O of, say, core-vulnerable or core melt states. Among these are a variety of issues in the arena of human factors, and these methodological issues stand apart from issues of human error quantification, about which we will say more below as part of our discussion of the error data base. A key example of methodological inadequacies is the failure of PRA methods generally to consider adequately the ability of reactor operators to cope, , through improvisation and ingenuity outside of standard procedures, with accident sequences as they develop in real time. The coping must surely O allow operators to terminate some sequences successfully that otherwise would develop further into high-risk mcidents. Yet we do not know how much difference this inadequacy makes to the final results. On the other .O O
O 24 side of the same coin is the possibility that even well-trained operators, O being fallible humans like the rest of us, might significantly aggravate a sequence that in the standard PRA analysis is terminated successfully without damage. O Another methodological limitation in a related arena is the present inability to model control system failures well enough. The O event-tree / fault-tree approach intrinsically views accidents through the concept of system functional failures caused by underlying component and support system failures or unavailabilities. In its present state of O maturity, the methodology can only incorporate control system failures through ad hoc analysis of when, and how, multiple component system failures might arise from control system failures. While this a_d hoc O analysis is probably adequate for most sequences, we do not know whether it does acceptably well when " time is of the essence," that is when rapidly developing events, especially in the early stages of some accident classes, O could severely tax the operators' capabilities or the resilience of some components. O Time sequencing issues also underlie uncertaint:les in event trec delineation (and, to a lesser degree, fault-tree delineation). The event trees are written down in a time-ordered fashion, but issues can rise of '3 "which failure occurs first" and of whether recovery might occur later in a > sequence. Again, whether this set of issues makes a significant difference j to the final results is not known, although our experience leads us to O expect that differences are unlikely to be large.
25 O O As described in the NRC's PRA Procedures Guide, quality assurance is very important in the generation and analysis of fault trees. For example, if the same event appearing in two or more places in a fault tree were O mis-typed when entered into a computer, one can make an error as large as a factor of 100 (usually non-conservatively) in computing accident , probabilities. (Such numerical errors would probably show up, of course, in O the final analysis.) It is important to note that we did not check sal's fault trees, nor did we run an independent computer analysis of their fault trees. O None of these trethodological limitations is special to the Shoreham PRA analysis by SA1: all are generally shared by other PRA studies on other O reactors. Indeed, progress is gradually improving our confidence that these limitations are not important enough to invalidate our confidence in the coilclusions of PRA . . . but they do temper our confidence, in an
- O unquantifiable way.
The tenor of the above discussion reveals our overall conclusion about the
- O methodology used by SAI in the Shoreham analysis. That conclusion is that the SAI methodology, though it suffers from some generic limitations comon to all reactor safety PRA's to date, nevertheless represents the D present state-of-the-art; it includes important methodological advances in some areas. We believe '. hat it is an acceptable basis for estimating the probabilities of the important internally-initiated accident sequences at 10 Shoreham.
O
O 26 3.2 Data Base Considerations , O The validity of the failure data base is vital to the validity of any PRA analysis. A couple of issues in our present case represent limitations O that might be importo.it. First and foremost, the Shoreham reactor is still in final stages of construction, so we have no data on failures at Shoreham: the failure data must generally come from industry-wide sources. O But operating experience has told us that some failure and unavailability rates can vary widely from plant to plant, and we don't know whether Shoreham will be above or below average. O The other side of this coin is that significant advances have been made recently in identifying below-average design, maintenance, and test O practices, by study of LERs and plant-specific attention to determining root causes of system and component failures. To the extent that these activities represent improvements, a new plant such as Shoreham can take O advantage of this experience to improve its performance over the average performance of plants already running. O Human error quantification is the other arena where data base issues seem to represent a limitation. The SAI study team has used the accepted industry-wide reference for most of its human factors failure data (Ref. O 4), but these are widely understood to contain large uncertainties. In particular, the failure data are lumped into broad categories whose applicability to specific sequence situations at Shoreham must surely be oaly approximate. Also, how Shoreham's operators will behave, c'ompared to industry averages, is of course not known. O
-:w ) 27 Despite the limitations just discussed, it is our conclusion that the ) Shoreham PRA analysis under review has used state-of-the art data bases generally. We have found some specific cases (see below) where we do not agree with certain specific numerical values, but we conclude that the data ) used are a generally acceptable basis for estimating accident probabilities at Shoreham. )
3.3 Specific Accident Sequences -- An Eclectic Critique To review quantitatively all of the important accident sequences that SAI quantified in its Shoreham analysis was not possible (and in our opinion not necessary) within the scope of this report. We took the approach of
). studying what SAI wrote down in their event-tree delineation, and using our experience and judgment to ascertain whether the approach and results were roughly congruent with our expectations. This is what characterized our accident-sequence review.
This activity consumed a reasonable fraction of all the effort spent on this project. In the course of it we found that most of what we studied was unarguably consistent with our experience and understanding, while ! there were only a few cases to the contrary (albeit important cases). For example, we have already mentioned above that we examined the list of internal accident initiators to ascertain whether SAI had left out any from l l tN list that we would have used, and we indicated that they had not ) I i
-.r .-m _ . , . _ _ _ _ . _ _ - _ . - - _ . - _
D 28 (except fires). We also attempted to look at some of the numerical values C that SAI used in order to see whether they agreed with our collective experience. In the course of ttis review, we found ourselves comfortable with almost all of the numbers we examined, the exceptions being noted J below. Perhaps the most important part of the PRA analysis where we find that our D review does not agree with the SAI draft report concerns their analysis of internal flooding (their section 3.4.4.1 and Appendix G). In particular, consider the situation in which portions of crucial safety systems are C disassembled for routine maintenance during plant operation. If during this disassembly the valved-off component is accidentally reconnected to its pipe (such as accidentally opening a motor-operated valve that has been 7 closed to allow the maintenance), then release of water through the opened valve will occur. If the mistake is not promptly corrected by re-closing the opened valve, an internal flood can result; such a flood at " Level 8" O inside the reactor building can quickly inundate several pieces of critical safety equipment, and its analysis as a special issue was deemed so important that an entire Appendix G is devoted to it by SAI. A detailed numerical analysis of this issue, by H. E. Lambert of our review team, is presented in Appendix D to this report. We will briefly discuss here the main issues and conclusion. We believe that SAI's analysis (3 contains some errors that underestimate the internal flood frequency, the correction of which would raise the core-vulnerable frequencies for Classes 3 I and II. O I
O 29 The events leading up to a disabling flood are as follows: O Event A: On-line maintenance of some critical system Event B: During maintenance, the system is disassembled Event C: Inadvertent opening of isolation valve, causing the flood Event D: Failure to reclose valve within specified time period O Event E: Operator erroneously isolates the power corsersion system during flooding causing the accident because the heat sink is lost We believe that the initiating event is Event C: the occurrence together O of Events A and B defines a vulnerable system state that permits C to initiate the f1 ad. Because Event C is an initiating event, we must compute its frequency, and also the failure on demand of Event D. g Event C's frequency (in units of events per year, or the like) must include the pre-existing presence of the vulnerability-inducing states A and B. i The units of the events, in the calculation, should be maintenance acts per n v year (Event A); probability of system disassembly given maintenance (Event B); conditional probability of inadvertent isolation valve opening given maintenance with disassembly (Event C); conditional probability of failure O on demand to reclose the opened valve (Event D); and conditional probability of erroneous operator isolation of the power conversion system (Event E), which then would initiate the accident. We believe that SAI has g incorrectly used system unavailability for Event A, and that their calculated result is about a factor of 100 lower than the value we obtain, all of which is explained in more detail in Appendix 0 to this report. g O o
0- 30 In addition, we believe that the SAI team has used an incorrect value for O the Event D failure (probability to reclose the accidentally open valve): SAI uses 0.05 for this probability whereas their primary reference source (Ref. 4) uses 0.25, if one accepts that there will be highly stressful O conditions during the period when the operator action will be required (see Appendix 0 for details).
- O An important assumption made by SAI is that flooding to the six-foot level will not ree"1t in automatic closure of the MSIV's. (SAI does assume, however, that reactor trip will occur.) It is important to verify that the O assumption regarding automatic MSIV closure is true. Otherwise the power conversion system is lost and the only normally available coolant makeup system is the condensate system. In this case the accident frequency O caused by flooding would increase by an additional factor of about 10, and a design change might be necessary to overcome the problem just discussed.
O If our flooding analysis is correct, then the internal flooding accident i frequencies are at least of the order of a few times 10-5 er reactor-year, and as described in our Appendix D could be much higher, depending O upon how human error rates are quantified. These accident sequences then become dominant for both Class I and Class 11 type accidents. O There is another part of the SAI analysis where our review team disagrees with the SAI work. Our difference of opinion enters critically into a key class of accident sequences, the so-called ATWS group ( Anticipated O O y, _, , ,m-,- ,w
O 31 Transients Without SCRAM), which comprise Class IV of the five classes into .O which SAI grouped the Shoreham. accidents. These sequences arise when one of several anticipated transients cannot be properly controlled because the
" SCRAM" system (which inserts the control rods) fails to ::iut down the O chain reaction. If this event were to occur, back-up engineered functions are brought into action to bring the reactor to safe shutdown: thes'e are discussed on pages 3-94 ff. of the SAI report and include the alternate rod O insertion system; the standby liquid control system to inject boron poison into the core; trip of the recirculation pumps; and operator procedures.
Of course, .ne critical failure that drives this sequence is the failure O to SCRAM on demand, for which there are essentially no empirical data upon which to base an analysis. O The SAI report selects as its value for failure to SCRAM on demand the likelihood 3 x 10-5 (about one failure in 33,000 demands). In its discussion (see p. B-111 ff), the report points out that "The calculation O of Scram system reliability has been an issue which has taken on both technical and philosophical aspects over the last seven years." This is because within the regulatory arena the issue of whether reactors are O adequately safe against ATWS events has been one of the most hotly-contested issues in recent years, whose regulatory resolution is still not complete. Unfortunately, there is also no consensus in the safety community about how to go about calculating the value for this failure nor about which value should be used in analyses such as the Shoreham case.
- O O
'O 32 While admitting that the situation requires a good deal of judgment, we O believe that tha value chosen by SAI is too high, by a factor in the range of 3 to 10 . . . that is, we would have chosen a value of 1 x 10-5 to 3 x 10-6 instead of their 3 x 10-5. This judgment, based upon nearly a D decade of study of the issue, cannot be defended analytically; but we do believe that the 3 x 10-5 value, which SAI seems to state (p. B-115) is chosen with a conservative bias, is too high. (Of course, we are aware that certain industry calculations fall in the range of 10-0 , or sometimes much smaller.)
O The impact of this judgment on SCRAM f ailure probability is practically linear in the results of core-vulnerable and core-melt frequency for Class IV. However, even though our best judgment would lower the likelihood of O Class IV accidents, we believe that it is prudent for Suffolk County to base its emergency planning on the Class IV core melt values as quoted in the SAI draft report . . . this approach takes cognizance of the observation that our difference of opinion with SAI is a purely judgmental one. 'O There were other areas that we reviewed which are important to the SAI analysi s . For example, as part of our review we studied the discussion on success criteria found on pp. 1-22 to 1-27 of the SAI draft report. It O is clear that in any analysis of accident sequences, whether the outcome is
" successful" (that is, leads to an adequately cooled core without c
vulnerability to core melt) depends crucially on whethcr the design lQ l iO
O 33 requires, say, all 4 low-pressure pumps or, say, only 1 out of 4 to cope O with a large LOCA. The success criteria chosen by the SAI analysis team have been taken, according to their report, from manufacturer's (GE's) reports and analyses, and the report comments (page 1-22) that "it is O believe that the success criteria so defined tend to be conservative." We were not able to assess this claim independently, although it makes some sense to us because the postulated conditions in the GE analysis are .O limited in a conservatively biased direction. We wish to call attention to this observation as a possible source of uncertainty in the SAI results, which if corrected would lower their calculated core melt valtes. O We also looked at the way human errors are incorporated into the component and system failures. Here our understanding is that the SAI team has O basically relied on the standard data source, the work by Swain and Guttmann (Ref.4), for the human failure data. Since this is the standard work and there is no better source at the present time, we have no quarrel
- C with the SAI team's decision here, but we feel it important to point out that there are very large uncertainties in the values for human failures that Swain and Guttmann quote. Even in the situations in which these O values are applicable, there are large variations in the error rates from one human to another, but the stickier problem is that the actual environmentsinwhichthehumanerrorswilloccurNeactoraccidentsarenot O those in wnich the error rates used are necessarily directly applicable.
Of course, this is a problem that is generic to reactor PRAs as a class, and not special to the Shoreham effort under review here. But again we O !O
34 wish to call attention to this issue as a source of possible high ) uncertainty in the SAI results. Certain other assumptions, necessary in the SAI analysis in most cases for ) want of any better approach, merit brief discussion. On page 1-30, for example, the SAI report says that "The failure of display of information to the operator is treated as a random independent failure or set of failures ) and is not dependent on the accident sequence." The limitatiJns in this assumption are clear: that is, independence is obviously not fully true for some sequences, but we understand (and concur with) the reasons why the , ) assumption is made. Also (p.1-28), "The maintenance contributions in the fault tree model are modeled as mutually exclusive among certain systems consistent with the Limiting Conditions of Operation. An example of this would be that HPCI is not allowed in maintenance if RCIC is unavailable." While this seems reasonable, it is clearly non-conservative to some degree, since LCOs are not always obeyed. Another guideline used in the SAI ) analysis (p.1-28) is that " Plant components are presumed to meet all performance requirements consistent with licensing." This assumption implies that, although a component or system might be unavailable, if it
) is operating it will perform essentially at least as well as the licensing requirement. Again, this reasonable approach contains some (unquantifiable) error. ) ) )
_ . _ . _ . . . . _ . . . . . . . . _ _ . _ _ . . . . J
O 35 3.4 Implications of the Review of Specific Accident Sequences O Our review of specific accident sequences described in the SAI draf t PRA report has resulted in several specific comments that have been discussed O in sections 3.1 - 3.3 just above. Here we will consider the implications of these findings for Suffolk County's emergency planning activities. O Our most important quantitative finding is an apparent error in the SAI calculation of the probability of severe internal flooding: we have discussed this in section 3.3 and in Appendix 0 of this report. If our O interpretation of the frequency of flooding at elevation 8 of the reactor building is correct, then the frequency of core-vulnerable conditions in accident Classes I and II is raised considerably: mostly in Class I in our O opinion, with some effect on Class II as well. The SAI report gives as its best-estimate for core-vulnerable frequency in Classes I and II the following: O Class 1 2.7 x 10-5 per year . Class II 1.1 x 10-5 per year O The conditional probabilities of core melt are given as 7% and 8% for these two classes, respectively. O If our analysis is correct, these core-vulnerable frequencies would be considerably higher. Because we have not been able to carry out a complete 3 systems analysis, we believe that our choosing a number and sticking to it 1 i t l
1 36 as "our value" is inappropriate: we believe it is more appropriate that
) LILC0 and SAI carry out a proper analysis. We not only believe that the core-vulnerable numbers will rise, but also tnat the conditional
,orobability of core melt might increase: to be precise, if flooding
) dominates Classes I and II, then the flooding sequences should be used as the basis for computing the conditional probabilities, with the possibility that core-melt might be more probable than 7-8% given the amount of
) equipment taken out of service by a flood at elevation 8.
With these provisos, we believe that it is prudent for Suffolk County to
) use a number in the range of 10-4 per year as the likelihood of core melt from both Class I and Class II: use of this value as input to the planning basis for the County emergency planning activities will prudently ) allow the County to protect against accidents at Shoreham within Classes I II. Also, because no core-vulnerable /: ore-melt analysis has been done for these internal flooding sequences, we believe it is prudent to ignore this factor for the time being in these classes. (As it turns out, using these higher values probably makes almost no difference to the Suffolk County emergency planning activities, because Classes I and II dominate the accident planning basis one way or the other.)
For the other accident Classes, the only one where we believe the SAI results are probably not correctly representative of the real accident frequency is Class IV (the ATWS accident group). However, as discussed ! earlier, even though our best estimate would lower the SAI values for l
37 C) core-vulnerable frequency by a factor of about 3 to 10, we recommend that Suffolk C ty use the SAI values as a prudent planning basis. O Our review revealed some important qualitative conclusions that should be me inad . Most importan', of these is that we have found the SAI C) ' analysis of at de t secuc .es and their probabilities to be on the whole a fully competent, state-of-the-art analysis. Also, we believe that t'here are major uncertainties in the results of these analyses today, not only the analyses by SAI but analyses by all teams that carry out reactor PRAs: we have touched on some of the underlying reasons for these in earlier discussion. Nevertheless, we believe that the SAI group has taken some O care to understand and quote the ranges of their uncertainties, and we find their uncertainty analysis for the accident sequences acceptable. O O O l l 10 a i
O 38 4.0 Are the Accident Phenomena Within the Plant Treated Correctly?
-r ')
The approach taken by our review team in reviewing the accident phenomena parts of the SAI draf t PRA was to ascertain whether the analytical metPods employed are reasonably consistent with the present state-of-the-art methods now used for these calculations. Members of the FRA team are quite familiar with the level of understanding within the broader reactor safety community about these phenomena, including a recognition of the significant limitations to our present understanding. n The basic conceptual problem being addressed here is as follows: we postulate the unlikely situation that the Shoreham reactor has reached a physical state in which a core-degradation accident is underway. The . accident sequence that has led to this physical state is assumed to be understood, in the sense that the event-tree /f ault-tree analysis has revealed to us a specific sequence of equipment failures and human errors leading up to the onset of core degradation. (It is not necessary for the purposes of this part of the analysis to know the probability of the sequence.) The problem is to attain acceptable understanding of the Q physical sequence of events following the equipment failures modeled in the event-tree /f ault-tree analysis, beginning with the onset of core degradation and continuing through until the reactor core is either O securely cooled, or has melted (wholly or partially), releasing a fraction of its fission products to the reactor vessel, the containment structures, and possibly the environment. O O
,0- 39 In order to carry out quantitative calculations of the phenomena, one O requires a variety of data, some conceptual understanding of the physical and chemical events, and a calculational model embodying this understanding. The calculational model should incorporate not only the
- O physical reality of the' reactor plant itself, but also the functioning (correct or degraded) of various engineered systems within the plant.
- O In an ideal analysis, it would be desirable to calculate the sequence of phenomena for every important accident sequence that follows core degradation. Unfortunately, though every accident sequence is different O from every other one, such a massive effort would be beyond the calculational ability of any team of analysts today. Also, because our understanding of thse phenomena is limited it would not make much sense to
;O model each specific sequence separately: differences among similar sequences are far less significant than uncertainties in our understanding.
O For these reasons, PRA analysts generally have grouped the numerous
- accident sequences beyond core degradation into categories, each category being treated separately in the calculation of core degradation / melt i
O phenomena. This grouping allows the analysts to treat a tractable set of , ' issues, with the limitation that errors and uncertainties are introduced I because of the approximate nature of the categorization process. ,
- O Each group of accidents is characterized by a set of conditions: examples include high or low pressure within the reactor vessel; containment t
'Q lO _ _ . ~ . . . _- . , _ . _ . , . _ . _ . . m_-b,--,_._ , _ , , - . _ , ,. . . . ~ . -
1
) 40 integrity intact or already failed before the onset of core degradation; ) time elapsed since SCRAM; various engineered safety systems either available or failed; and so on. ) The challenge is to develop an acceptably accurate quantitative analysis of what happens to the core and its fission products, and how, and in what sequence, and under what phyhsical-chemical conditions. The result of the ) analysis should be a quantification of fission product releases from the plant in terms of species, quantities, times of release, physical-chemical conditions of the release (energy of the release, physical form, etc.), and ) some measure of the confidence of one's conclusions.
FRA's purpose in this review has been to develop an understanding of what
) the SAI analysis team has done in the course of their work, so as to ascertain whether their results form an acceptable basis for the calculation of offsite consequences: if the SAI calculations do form such ) an acceptable basis, then Suffolk County's emergency response plan can utilize them.
The approach taken by the SAI analysis team was to utilize a well-known, widely used, and well-documented set of computer-based models known as the l l MARCH-CORRAL code package. These codes, originally developed under AEC/NRC sponsorship for the 1973-74 WASH-1400 analysis, have been the main method of analysis of these phenomena ever since. However, it is widely known f that these codes contain important approximation's, omissions, and other ) T- - v- --- ,n ---- --
41 O limitations that make their modeling of physical reality less than precisely correct, and indeed the limitations of MARCH and CORRAL have been g a continuing subject of research, analysis, and discussion in the reactor safety community for much of the last decade. In addition to the MARCH-CORRAL codes, the calculations of the accident phenomena require O various types of input data, the most important of which are the fission-product release fractions or partition fractions in various stages f the postulated accidents. Examples include information about-fission O product releases from the fuel pins during fuel melting; fission product partitioning in a water / steam environment as a function of pressure, temperature, and ther conditions; and fission product plateout, transport, O and the like in the airborne state. The MARCH-CORRAL code requires data such as these essentially as parametric input to its modeled calculations. O The SAI analysis team made modifications to parts of MARCH and CORRAL in order to tailor the code to the Shoreham reactor, and to incorporate some recent insights about various phenomena. These modifications do not O substantially modify the basic operating philosophy of the codes, nor their major limitations. O 4.1 Grouping of Accidents into Classes O One important question that we have looked at is whether the grouping of accidents into five classes by the SAI team is reasonable. The five Classes can be briefly characterized as: g 0
) 42 Class I Loss of Coolant Makeup Accidents Class II Loss of Containment Heat Removal
) Accidents Class III Large LOCAs Class IV ATWS (Anticipated Transients Without Scram) without Poison Injection Class V LOCAs Outside Containment (including Interfacing Systems Accidents) ) All accidents leading to and beyond core-vulnerable conditions are placed in one of these five Classes, and the five Classes are dealt with ) separately in the subsequent (MARCH-CORRAL type) analysis. We have studied the validity of this approximation, and find the grouping by SAI to be fully acceptable. We believe that some differences exist among accidents ) that are grouped by SAI into a common Class, an example being the varicty of accidents grouped under Class II in which containment residual heat removal f ails: the very long time that 't takes for the Class II accidents ) to evolve is characteristic of the Class as a whole, but differences in detail emerge depending on which safety systems fail in which order. However, in this case (as generally), we believe that the grouping is a ) reasonable approach. Our reasoning is that the uncertainties in the phenomena are sufficiently great as to overwhelm any additional "information" that might have been gained by a grouping into more numerous ) Classes; and that the choice of specific sequences within each Class to serve as the model for each Class is also reasonable. b An important issue is the time duration between initiation of the accidents and the eventual release of fission products to the environment: this will differ from Class to Class substantially. We have studied the general ?
O 43 grouping from this perspective, because of a concern that perhaps there O could be mis-classification from the perspective of time evolution. We tend to affirm that the time evolution of the Classes is reasonably consistent with our experience: for examp".e, the Class II accidents evolve O very slowly compared to the other Classes, and we believe that all of the accidents grouped into Class II behave in this way. This conclusion allows us to deal with all of these accidents together in discussing their O emergency-planning implications in the context of Suffolk County's needs. O 4.2 Review of Phenomena After Meltdown The approa.h we have taken for ascertaining whether the SAI meltdown C analysis of each accident Class seemed reasonable was in part the use of subjective judgment based on the experience and knowledge of the reviewers, and in part some independent calculations of a few of the phenomena O involved. These independent calculations, carried out by Dr. Fabic, tend to support the validity of the SAI conclusions for many of the important post-meltdown phenomena, with some exceptions that will be discussed next. O A general elevation-view drawing of the Shoreham reactor and containment building is shown on the next page, reproduced directly from the NRC Safety Evaluation Report for Shoreham (Ref. 6). The main important features to O D
44
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J concentrate on for our discussion below are the location of the suppression pool at the bottom of the building, and the downcomers penetrating the drywell floor below the reactor itself. That the MARCH code (which models the highly complex processes involved in a meltdown) has deficiencies is, as mentioned above, well known. In Appendix B, some of these are discussed by Dr. Fabic, and his discussion will be excerpted here, as follows:
"In the course of the SASA (Severe Accident Sequence Analysis) calculations being performed by the Oak Ridge National Laboratory (ORNL) for the USNRC as applied to BWRs, the ORNL engineers have also come to the conclusion that MARCH has many deficiencies. Out of 21 listed inadequacies, we shall extract only a few:
- Modeling of heat transfer to upper and lower BWR vessel internals
- Modeling of core collapse
- Failure of bottom head via control rod drive tube penetration not considered
- Suppression pool and wetwell/drywell interaction
- Rod-to-rod radiation heat transfer not included
- Vessel water level calculation does not include variable flow G
areas
- Fuel pin melt / slump / freeze phenomena are not mechanistically modeled."
9 9
)
46 l Dr. Fabic (Appendix B) also discussed the reasons why the MARCH code I contains extensive simplifications:
)
"The extensive simplifications in the code were introduced for two reascns: (1) to produce a tool useful for probabilistic risk assessment which requires many computer runs for exploration of
)
consequences of various bounding assumptions. Hence, computational economy must have been one of the principal goals; (2) to produce a tool amenable to accepting some of the major uncertainties as input
)'
assumptions that could be changed from run to run. These stem from the relatively poor knowledge of various thermohydraulic processes that involve melt propagation and the attendant heat and mass
)
transfer in complex geometries. The pertinent empirical base lags far behind the existing empirical base collected in the course of reactor safety research for situations that do not involve a
)
degraded core. However, the word ' uncertainty' is also used here to imply simplifications one needs to make to intentionally bypass the detailed calculations that would cause the code to become long ) i running." Despite these limitations within MARC 4, the conclusion of the FRA review
)
team is that the code has been applied by the SAI team in a competent and conscientious manner. Our analysis of their description of their MARCH runs, the way they have thought about issues involved in MARCH, and their
)
O 47 conclusions has left us with both a confident feeling about what the SAI O analysts have done and a heightened appreciation of the approximations inherent in the MARCH modeling. O One of the important areas of disagreement concerns the mechanism of melt-through of the molten core as it penetrates the bottom head of the reactor pressure vessel (RPV). Again we quote from Appendix B, Dr. Fabic's ~ report (the abbreviation "CRD" means " control rod drives"):
"We do not believe that the RPV bottom head would fail in either of O the two ways described in the Shoreham PRA: structural failure of the vessel wall due to elevated pressure and temperature or the wall melt-through when the RPV pressure is low.
- O "It tead, we agree with the scenario described by R. Henry of Fauske Associates, Inc. (see item 8 in Section 1.1) wherein the relatively thin metal that seals the CRD tube, or the CRD tube itself, is much more likely to fail first upon contact with the melt.
O "The melt is very unlikely to reach the bottom head in a coherent fashion (gross slumping). It is more plausible to consider downward streaming of melt around the vessel axis. That melt will not attack the CRD tube as long as there is some water left in the lower plenum. If the amount of steam generation caused by quenching of melt is insufficient to stop further melting (a likely case), 'O .O
O 48 increased amount of debris will accumulate on the bottom portion of O the RPV and it would eventually remelt the fraction of debris that was frozen by the liquid--which by now has evaporated. This whole process could be delayed significantly if the CRD cooling water were O continuously supplied.
" Eventually, one or more CRD tubes would fail and the debris !C discharge into the CRD room would commence."
Following the failure of the CRD tubes, another key question is how fast
- O the melt will pour or stream through these CRD channels. Dr. Fabic calculates (Appendix B) that under the conditions likely to be present, only a very few seconds (perhaps less than 5 seconds) would elapse before O as much as 50% of the melted metal would flow out of the lower vessel openings. In this scenario, there is no outright structural failure of the vessel wall, nor wall melt-through per se. However, as the melt O penetrates the CRD tubes, significant enlargement of these penetrations will occur by ablation. If this is true, then only a small fraction of the
- melted core will remain in the bottom head, and that fraction will be the O core material that has not melted, or has re-frozen at the bottom. Whether the flow thrcagh the CRD openings is pressure-driven or gravity-driven will make a difference, but should not make a very large difference to the O phenomena in this process, which will be quite fast.
l lO I { 1
) 49 The next issue involves where the core debris streaming through the CRD ) openings will go. With a large amount of CRD hardware in the way (essentially hanging down in large massive metallic drive channels below the lower head), and with such a fast outflow, it seems unlikely that the ) CRD hardware will melt before the core material has passed through, which in turn indicates that the core material will reach the drywell floor only after being significantly broken up and " sprayed about" by the CRD ) hardware. However, an important insight (see Appendix B) is the conclusion that most of the core will go down into the drywell area just below the reactor pressure vessel (RPV) head, with rather little of it going into the ) annular outside region of the drywell floor. Once the level of core melt debris on the drywell floor exceeds the height above the floor (about 3.5 inches) of the downtomer top flanges, the molten debris will start pouring ) down the four downcomers that are located within the area of drywell floor just below the RPV: we find that it is unlikely that much of the melt will go down the numerous downcomer channels in the outer annular region, since ) the melt will go down these four inner downcomers promptly. Our analysis indicates that this will occur rather quickly, most likely within about 5 minutes after RPV breach. ) Once the debris begins pouring down the downtomers into the deep wetwell, we studied whether steam produced by the molten material would produce ) enough countercurrent flow upward to impede further melt from penetrating downward. Our calculations (see Appendix B) seem to indicate that this phenomenon will not occur. ) )
.O 50 In the scenario described above, almost all of the melt will end up in the C wetwell almost all of the time. Appendix B discusses why we are reasonably confident of this conclusion. It appears that there is only a limited range of conditions where melt will not mostly be quenched in the wetwell.
O The implications of this quenching are, of course, that significant removal of fission products in the wetwell water would occur, lowering by large factors the amount of fission products other than noble gases available in O the gaseous phase for atmospheric release through an ultimate containment breach. D Perhaps the most important conclusions from Dr. Fabic's analysis in Appendix B are that vessel failure will probably take place through the CRD channels; that discharge from the vessel to the drywell will be very rapid O (less than 1/2 minute), and discharge from the drywell to the suppression pool will also be fast (less than, say,10 minutes); that core melt material will nearly always end up in the wetwell, where significant 1 0 removal of fission products will occur; and that there is little time for debris-concrete interaction on the drywell floor, and too low a temperature l for that interaction to-occur on the wetwell floor. 'O All in all, the picture painted by our analysis is that the SAI results on core-melt behavior and fission-product release are likely to be O conservative (that is, too high values for release, too much core-concrete interaction) compared to what we believe to be the real phenomena. O
- O
i l
) 51 l
In particular, we expect that for two of the classes of accidents (Class I*
) and Class III**), there will be significant delay in the containment rupture time, and perhaps no containment rupture at all, depending on specific containment heat-leakage properties that we have not specifically ) addressed and that are both complicated and scenario-specific. Also, there will be quite large decontamination factors for fission products released in the wetwell pool, leaving the gaseous fission products released directly ) from the melted fuel pins as the principal airborne species available for release from the containment after breach. ) Unfortunately, we are not able to quantify the extent of this conservatism.
The phenomena that we believe will occur differ from those modeled in the SAI draf t report in several areas, but almost always in the " conservative direction," meaning that if our analysis is a better representation, then the accidents will be less severe. We believe that much remains to be done before a reliable calculation of these phenomena can be carried out: some research on specific phenomena must be performed; some improved modeling of the sequence and character of the events must be accomplished; and some i better data on specific coefficients must be obtained, for a variety of physical and chemical interactions.
- Class I involves accidents where failure of core cooling with the RPV at high pressure af ter a transient or small-break LOCA results in core degradation, with an intact containment until after core melt.
** Class Ill involves accidents such as a large-break LOCA inside containment where RPV depressurization occurs prior to cora degradation, with an intact containment until after core melt.
l
) 52 4.3 Some Specific Issues Involving Meltdown and Related ) Phenomena In the course of our reviGw c. the meltdown part of the draft PRA on ) Shoreham, a number of specific issues arose that we believe are worth discussing within this report. These are technical issues that could affect the overall " bottom line" conclusions of the SAI study. For each we ) will indicate what our own conclusions are as to the impact of the issue vis-a-vis Suffolk County's application of the PRA results for emergency response planning.
The first issue is the magnitude of fission product releases from the fuel. In Table D-3 (page D-9) of the SAI draf t report, a list is given of release fractions from the fuel to the primary reactor system during core meltdown. The table is reproduced on the next page. Shown in the table are two values for the release fraction for each isotope: one is the value used in k the 1973-74 WASH-1400 analysis (Ref.1), and the other is the more recent value found in NRC report NUREG-0772 (Ref. 5), which contains a discussion of recent understanding of fission product behavior in core-melt accidents. The SAI report states (p. D-7) that they have used an average of the two model estimates (WASH-1400 vs. NUREG-0772), in view of uncertainties in our knowledge. While this averaging does not make very much difference to most of the release values, the value for tellurium is quite different in the two models: WASH-1400 assumed that only 15% of the tellurium escapes from the fuel, while NUREG-0772 believes the correct value is 100%. While we do
) ,
53 fuced directly from the Shoreham PRA by SAI) Table D.3
)
RELEASE FR< TO REACTOR PRIMARY SYSTEM DURING CORE PELTDOWN
> Fission Group RSS MUREG-0772 Xe, Kr .9 1.0 I, Br .9 1.0 Cs, Rb .81 1.0 Te .15 1.0 .
Sr .10 .3 ) Ba .10 .5 , u** .03 .02 Nt .003 .03 ) g, .003 .003 r l l
- Inc 2 Sb, Se l ** Inc s Mo , Pd, Rh , Tc
- Inc s Nd, Eu, Y Ce, Pr, Pm, Sm. Np,~Pu f
NOTE: '5" is " Reactor Safety Study", Report WASH-1400 l I 1 ) 5 ? l t
) 54 not know which is correct (that awaits further experiments that have not ) yet been done), we believe af ter study of the NUREG-0772 arguments that the lower figure is more likely to be correct for tellurium, which would slightly decrease the offsite releases and doses. The effect is a small ) one generally, but not so small as to be negligible. SAI's draft analysis also states (page D-7) that 90% of the iodine is } released from the primary system as Csl (cesium iodide), which is a major departure from the WASH-1400 assumption that iodine was released 100% in elemental form. This makes a significant difference because the Csl is ) soluble while the elemental iodine, in gaseous form, was a key contributor to offsite doses in WASH-1400. We tend to agree with the recent arguments-that elemental iodine is not a likely form for that element in these ) accidents, and believe that SAI's assumption is a reasonable one. However, there are still important uncertainties in our knowledge of what precisely happens to the iodine during the accidents under consideration. The ) reactor safety community will require some experiments, some modeling, and considerable discussion (all of which is underway) before this important issue is resolved and a consensus reached on it. ) Another issue involves whether there would be zirconium-water reactions within the reactor vessel af ter the fuel has melted but before the melt ). 541 assumed (page C-23) that in a melted state penstrates the bottom head: the Zr-water reaction would not occur, and has explained their assumption (Ref. 2) as being due to the presence of a molten pool at the bottom of the )
O 55 lower vessel head, covered by a crust at the interface between molten and
.O solid fuel. This crust would inhibit Zr-water interactions, according to the SAI opinion. While this scenario seems reasonable, the generation of zero Zr-water reaction seems too extreme, because surely some of the O zirconium will be in contact with water or steam during parts of the meltdown process. This issue needs more analysis, in our view. We have no way of quantifying whether the Zr-water reaction is miniscule or only 'O "small," but we do concur that it is unlikely to be "large* in the sense of comprising any reasonable fraction of the total zirconium in the fuel cladding. Thus we do not think this issue should have any important impact - .O on the overall results of SAI's analysis.
Another issue that we have studied is whether there will be rapid O degradation of the concrete walls of the drywell during core-melt accidents, which might lead to carbon-dioxide evolution, high pressures building up inside, and uR5 ate pressure failures of containment. First,
- O we should point out that in our own analysis, almost all of the melt goes directly from the bottom vessel head to the CRD room floor to the i
suppression pool; almost none goes to the annular drywell floor (see O Appendix B). However, for this discussion we will assume that we are considering melted core on the drywell floor. When our concern with the issue of drywell concrete degredation was raised with the SAI analysis l C team, their response (Ref. 2) was that only for the Class III ser;uences would a sufficiently high temperature occur in the drywell, ar.d these sequences produced very short thermal transients. If this is true, the O O l
56
- O decomposition temperature would be reached only for a very thin layer near
.O the surface of the concrete. We have not been able ourselves to calculate any sufficiently detailed numerical values for the temperature transients, nor can we find them in the draft SAI report. However, based on our O experience with this issue, and our understanding of the phenomena that will occur during melting and core transport from vessel to drywell to wetwell, we believe that this concern is not likely to be a major ene: .
LO specifically, we believe that even if significant core melt were to reach the drywell annulus, it is quite unlikely that this concrete-degradation effect will produce enough non-condensible gas to challenge containment O earlier than it is challenged in the scenarios set down by SAI in their report. O We have also been concerned that containment integrity might be compromised by pre-existing containment leaks, that might not have been considered properly in the SAI analysis. Also, we were concerned with failures of O penetrations at high temperatures. While SAI confirmed (Ref. 2) that this latter problem was not explicitly considered in the PRA, they believe that the former is adequately analyzed within the containment event trees. We aie not convinced that the analysis incorporates either of these issues h quantitatively; however, we believe that containment leakage is not as l l important an issue for the Shoreham BWR as it appears to be for many other O reactors (in particular, for many of the PWRs), because the effective fission product removal action of the suppression pool provides such large ! reductions for the most likely accident classes. Thus the issue probably does not introduce major additional pathways to the environment oven if it has not been thoroughly considered in the SAI analysis. !O-
) 57 > ) One of the key issues that has been an outstanding concern for many years in consideration of core-melt accidents is " debris-bed coolability," which is a shorthand phrase for the question of whether, if a bed of core rubble ) were to form from molten-then-solidified core material, that bed can be cooled adequately. Another question is whether steam, generated in and around the debris bed as it is cooled, can provide enough additional ) pressure to pose a challenge to containment.- First, we must clarify that we are concerned here with cooling a debris bed on the drywell floor; we have no concern with cooling the later debris bed under many feet of suppression ) pool water, although ultimately af ter many days the pool water would boil away and the issue would arise again. The SAI response to our inquiry on this subject (see Appendix C) is reasonable, and is congruent with our own ) judgment on the matter. However, it must be emphasized that while our judgment and that of the SAI group agree, the entire subject of debris beds is still one where everybody's conclusions are highly speculative; in our ) view the safety community as a whole doesn't have enough experimental data, nor modeling talent, to put this issue to rest at this time. Fortunately, the issue is not believed by our group to pose significant additional risk ) potential, for the following reasons which are quoted here from Mr. Davis' discussion in Appendix C: )
"Upon further consideration of this issue, it is considered not to be significant in terms of the potential for risk increase for the following reasons:
) )
O- ' 58 "1.If containment failure has not occurred at the time debris bed O coeling occurs, most of the fission products (which are released dut 1g the initial meltdown phase) will be securely trapped in the suppression pool water. Thus, the possibility of containment O failur- ' rom a steam pressure surge upon debris bed cooling would not c< ea large fission product release. O 2.If containment failure occurs before debris bed cooling, the major consequences of the accident would be underway, and the added fission product release would likely not be significant. O "3. A steam pressure surge sufficient to challenge containment integrity requires a large amount of water delivered to the bed O and intimate mixing. It is not likely that a high volume water source would be available since most such sources within the containment must previously have been asssumed failed or degraded to cause the accident to progress through core meltdown." 4.4 Core Vulnerability vs. Core Melt: Containment Event Trees O In section 3.1 (above), we commentea favorably on the methodological advance employed by SAI in the Shoreham study in which an accident that O proceeds to a " core-vulnerable" state is differentiated from one that proceeds further to a " core melt" state. As discussed earlier, this differentiation was not made in the WASH-1400 analysis, and yet it is O O
,0 59 clearly important to recognize that only a fraction of " core-vulnerable" O accidents proceed to " core melt," an example of one that stopped short being the Three Mile Island accident of 1979.
'O The SAI report finds, for the five Classes of accidents considered, the following conditional probability that a core melt will follow: Conditional Probability of Melt, Given Class Core Vulnerable State O I 8x 11 7% III 84% IV 43% V 100%
- O The way these conditional probabilities were calculated in the Shoreham PRA analysis represents an advance over earlier probabilistic reactor analyses.
The effort consisted of the development of very complex containment event O trees (CETs) that considered the large variety of engineered safety features and phenomena that are brought into play during the time period after a core-vulnerable condition is reached. Of course, some assumptions O are necessary to simplify the problem, and the SAI analysis team made several of them. We have studied them in the SAI draft report (see the ! list on p. 1-19, pages H-11 to H-13, H-48 to H-56, and Table H.7), and find
- O them generally reasonable, although the analytical basis for the specific i
numbers is weak. l C We believe that there are major uncertainties in the CET analysis, but our study of the SAI discussion leads us to believe tnat the SAI analysts were aware of these and handled them acceptably. For example, a key limitation O is that there are essentially no data on the relative likelihoods of containment f ailure modes of different sizes and types that would result lO
) 60 overpressure (see page H-50). Another example is that, although the ) analysis correctly differentiates between containment overpressure failures near the top as opposed to near the bottom of the containment (a failure near the bottom, although very unlikely, could compromise the suppression ) pool), their treatment is obviously an approximate one. We believe that the point of departure of the SAI treatment is proper. They start by differentiating among three situations: those in which the core is vulnerable to melting in an intact containment; those in which containment may be vulnerable first while the core is still adequately ) cooled; and those involving containment Dypass. These three topologically different states are then treated separately in the quantification process. We also endorse the approach taken in the draft SAI report to consolidation of the release end-states. This consolidation has the effect of grouping numerous different accidents and treating each group singly, which inevitably implies loss of detail in the interest of calculational tractability. The SAI report acknowledges this issue (p. H-51), and states that the approximations made are conservative in nature. We affirm that this is probably correct but have not reviewed the details sufficiently to have an independent view of whether this is always true (see Table H.8,
- p. H-106 for details).
) To mention other technical issues, we believe that the treatment of the steam explosion issue is a reasonable one. Also, the discussion on ) )
G1 o quantification of the CETs recognizes explicitly some concerns that clearly affect the quality of the results (see Table H.6, p. H-57 to H-60). The O SAI analysts have explicitly differentiated among various qualities of their supporting data, have given not only their best point estimates but 10%-90% bounds for their results, and have documented their main O assumptions. Again, however, as elsewhere in this review there was no that our review effort could examine the (literally, hundreds of) detailed numbers in the CET quantification. O Our failure to review these conditional probabilities quantitatively is not very troublesome to us, because we are of the opinion (a qualitative O opinion, however) that the velues quoted are reasonably within the range that we would expect. Furthermore, we believe that the inclusion or exclusion of the SAI conditior,al probability factors does not make any O important difference to what Saffolk County will do in its emergency planning activities. Therefore, and because we are uncertain as to the quantitative validity of the results, we believe that it is prudent for the O County to ignore these factors in its planning, and to take the core-vulnerable figures, as modified by our recommendations for Classes I and l II, as the planning basis for emergency preparedness. ,o Our rationale for recommending the exclusion of these factors is as follows: for Classes I and II, we have already recommended that the h County assume that the core-melt values lie in the range of about 10-4/ year, taking into account our improved analysis of internal l flooding; for the internal-flooding sequences there does not exist a valid O O
!O 62 core-vulnerable / core-melt analysis, and we believe that our recommended O r. umbers are in any event only a rough estimate. For Classes III and V, the SAI factors are so close to unity (84% and 100%, respectively) that there is no difference. For Class IV, the SAI factor is 43%, within O about a factor of 2 of unity, but for this class we believe that the core-vulnerable number could be too high by a factor of 3 to 10 because of SAI's use of too high a value for failure of SCRAM on demand: so for Class lC IV the inclusion or exclusion of a 43% factor is practically like splitting hairs.
- O For all these reasons, we believe that omitting the core-vulnerable /
core-melt factors frora Suffolk County's planning basis is the prudent choice: the factors have no reasonable basis for the internal-flooding
'O sequences and make essentially no difference in the other Classes.
4.5 Implications of the Review of In-Plant Phenomena O Our review of in-plant accident phenomena described in the SAI draft PRA report has resulted in a collection of specific comments and remarks that O are covered in the earlier sections (4.1 - 4.4) of this chapter. It is important to describe the context in which these comments are to be understood. We believe that the present state-of-the-art of probabilistic 'O analysis of in-plant phenomena is not very well advanced. In particular, our underlying understanding is inadequate for some phenomena (for
- O i
!O
D 63 example, core melting itself, core penetration of the vessel, debris bed O formation and coolability, aerosol plateout, core-concrete interactions); for the performance of some key systems (for example, containment failure mechanisms, effectiveness of suppression pool heat removal O mechanisms, efficacy of active aerosol removal systems); and for the time sequence and duration of some events during the accident (for example, duration of meltdown itself, pressure buildup). Given the paucity of experimental information, only limited applicability of the data that do exist, and calculational intractability of models O complex enough to contain detailed differential effects, it is no surprise that differences of opinion exist within the professional community. Some of these dif ferences are reflected in our comments above. What is O important to leave with the reader of this report is that we do not believe that there are important differences between what SAI has accomplished and documented and our own view about release magnitudes: differences of interpretation, differences in level of detail, differences in selecting experimental data or modeling approximations seem in every case to produce effects on the final PRA results that are within the quoted uncertainties of the SAI analysis. If there is one overriding impression that our review team is left with in the aftermath of this in-plant phenomena review effort, it is that the magnitudes of the radioactive releases are likely, in actual accidents, to O O w
O 64 be smaller than are calculated in the SAI draft report. This is due to O the inherent introduction of numerous conservatisms in the analysis. Paramount among these are conservatisms in the description of removal mechanisms that will keep important fission products within the reactor O building, for which incomplete credit has been taken in the analysis. Examples include the likelihood that bottom-head penetration may not even occur at all for some scenarios, either because melting will be incomplete. O or heat transfer larger, or CRD cooling water available (see our Appendix B); the analysis of suppression pool decontamination factors (see our Appendix C); and the various assumptions that SAI has made on containment .O failure mechanisms themselves. 1 We are unable to quantify the extent of these cor.servatisms in the in-plant O phenomena analysis; indeed, we believe that it will be several years from now before enough research has been carried out to enable a consensus to be reached on these issues. Fcr this reason, we believe it imprudent to take 'O account of them for the purposes of advising Suffolk County's emergency preparedness effort. However, it is important that the reader understand our view that the SAI radiation release results, taken at face value and ,O considering their large quoted uncertainties, are more likely to be too high (that is, to represent accidents more severe than actual) than too low. O O O e- .ee. -,
O 65 5.0 Summary and
Conclusions:
Are the Probabilities and
- O Magnitudes of_the Calculated Radiation Releases Correct?
In earlier sections we have already given our sumary conclusions as to the O correctness of the probabilities of the accident sequences (Section 3.4) and the magnitudes of the calculated releases (Section 4.5). Here we will sumarize our findings, and discuss their implications when applied to 'O Suffolk County's emergency planning activities. Concerning the SAI ca'culation of accident probabilities, we have found O two important differences between our analysis and the SAI analysis. They were noted in Section 3.4, as follows: First, we believe that due to an error in the SAI treatment of internal flooding, the contribution of O internal flooding to Class I and II accidents has been underestimated. While we are unable to provide our own analysis of the internal flooding accidents, we recommend that Suffolk County use, as a basis for their O emergency planning effort, values of about 13-4 per reactor year for the frequency of core melt for both Classes I and II. Second, we believe that Class IV accidents will occur less frequently than the SAI report claims, O because we believe the SAI report has used too high a value for the failure on demand of the SCRAM system. However, we recommend that the County should use the SAI results for Class IV as their planning basis. We O recommend that the County use all other values for core-vulnerable and core-melt frequencies directly as found in the SAI draf t report. O -O
66 Concerning the SAI calculation of magnitudes of radiation releases from v Shoreham, our review has laf t us with the strong impression, which is supported by the discussion in our Chapter 4, that the releases calculated in SAI's study are more likely to be too high than too low (that is, we O believe that their calculation has conservatisms within it). We have been unable to quantify the magnitude of these conservatisms, and in fact we do not believe that it is possible to do so at the present time, because of C insufficient information about the phenomena that characterize accidents of the type under discussion. However, we believe that the conservatisms could amount to large factors of reduction for some accident types, O particularly those important accidents (albeit, in an absoluto sense quite rare accidents) where the melted core would pass into the Shoreham pressure suppression pool. In our opinion the suppression pool, and the way the
' Shoreham design provides for prompt passage of melted core material to it in these unlikely accidents, will be very effective in fission-product remov al . The substance of our remarks here is our opinion that the removal O effectiveness of the pool may be even greater than the SAI analysts have used in their calculations.
- Thus we arrive at the following overview of our conclusions as to the probabilities and magnitudes of SAI's calculated radiation releases from accidents at Shoreham. We believe that the Class I and II accident groups
# have higher probabilities than are found in the SAI report, because of internal flooding accidents; and we believe that the magnitudes of the releases are likely to be lower than SAI has calculated.
3 0
O 67 The implications for Su' folk County's emergen:y planning effort will depend O upon what the County planners decio.1 to do with our findings. We believe that the implications for the County of higher core-melt probabilities in Classes I and II would be minor because these classes already dominate the 'a' composite offsite doses from Shoreham, sen if SAI's lower numbers are used. If the County planners use the composite dose-distance curves as their planning basis, increasing the absolute probabilities of these curves O changes nothing that the County will do or decide. Concerning the magnitudes of the releases, use of substantially smaller releases would have a major effect on the County's planning effort; however, we recommend O that the County utilize the SAI draf t results as their planning basis because we cannot quantify the degree of conservatism in them. O O O O O O
[) - 63 O 6.0 ACKNOWLEDGMENTS 1 The authors wish to acknowledge the advice and assistance of Fred C. O Finlayson, our collaborator on this project without whose help the work could not have been accomplished. Fred has also served in a liaison capacity with Suffolk County throughout this work. We also acknowledge O Christopher McMurray, counsel to Suffolk County, for much assistance. We have benefitted from important interactions with Robert Erdmann and Edward Burns of Science Applicctions, Inc. and Voyin Joksimovich of NUS 4D Corporation, who have explained parts of the SAI report to us. Finally, our interactions with Suffolk County itself, through Frank Jones and Patricia Dempsey, have been smooth and successful throughout. O O O 10 !O 1 l C m .--% .vt .-e--- - - - - - - n--- - - --, . - v---
i l I O 69 7.0 References O The primary reference used in this report is the probabilistic risk assessment carried out by Science Applications, Inc. on the Shoreham Plant. The reference is: O Science Applications, Inc. (San Jose, California),
"Probabilistic Risk Assessment, Shoreham Nuclear Power Station, Long Island Lighting Company," Preliminary Draft Report SAI-001-83-SJ, March 1982 The foHowing are referred to by number in the text:
O
- 1. U. S. Nuclear Regulatory Commission, " Reactor Safety Study,"
Report WASH-1400, October 1975
- 2. Letter, E. T. Burns (of Science Applications) to R. J. Budnitz (of Future Resources Associates), dated 2 August 1982, on the O subject " Responses to questions generated during July 16, 1982 meeting"
- 3. "PRA Procedures Guide, A Guide to the Pe.-formance of Probabilistic Risk Assessments for Nuclear Power Plants,"
prepared by the American Nuclear Society and the Institute of O Electrical and Electronic Engineers, Report NUREG/CR-2300, Review Draft, dated April 5, 1982
- 4. A. D. Swain and H. E. Guttmann, " Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications,"
Report NUREG/CR-1278, April 1980 g
- 5. U. S. Nuclear Regulatory Commission, " Technical Bases for Estimating Fission Product Behavior During LWR Accidents,"
Report NUREG-0772, June 1981
- 6. U. S. Nuclear Regulatory Commission, " Safety Evaluation Report n" Related to the Operation of Shoreham Nuclear Power Station, Unit No. 1," Report NUREG-0420, April 1981 O
l O t ,0
) 0 > g REVIEW AND CRITIQUE OF PREVIOUS PROBABILISTIC ACCIDENT ASSESSMENTS FOR ) THE SIIOREllAM NUCLEAR POWER STATION Report on Task I of the Project
" Consequence Assessment for Suffolk County Radios ,gical Emergency Response Plan" VOLUME II: APPENDICES Prepared for Radiological Emergem y Response Plan Steering Con.inittee
) Frank Jones, Chairman Office of the County Executive County of Suffolk II. Lee Dennison Building IIcuppauge, New York 11788 Robert J. Budnitz Principal Author ) Peter R. Davis Stan Fabic Iloward E. Lambert Contributing Authors ) September I?,1982 ) . ) FutureReaurces Amiates,Inc. ] 2000 Center Street Suite 418 Berkeley, CA 94704 415-526-5111
D . Oo j REVIEW AND CRITIQUE OF PREVIOUS PROBABILISTIC ACCIDENT ASSESSMENTS FOR 9 TIIE SilOREllAM NUCLEAR POWER STATION Report on Task I of the Project
" Consequence Assessment for Suffolk County Radiological Emergency Response Plan" O VOLUME II: APPENDICES Prepared for Radiologice' Emergency Response Plan Steering Committee 3
Frank Jones, Chairman Office of the County Executive County of Suffolk II. Lee Dennison Building IIauppauge, New York 11788 Robert J. Budnitz Principal Author O Peter R. Davis Stan Fabic Iloward E. Lambert Contributing Authors O September 17,1982 o . o Future Resources A.uociates,Inc. o 2000 Cen ter Street Suite 118 13erkeley. CA 9-170-1 -115-520-5111
)
TABLE OF CONTENTS, VOLUME II
)
, APPENDIX A Excerpts from the Contract between Suffolk County and Future Resources Associates, Inc.
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APPENDIX B Review of Within-Containment Phenomena and Containment Challenges APPENDIX C A Selected Review of the Probabilistic Risk Assessment Prepared for the y Shoreham Nuclear Power Station APPENDIX D Internal Flood Analysis
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D 0 APPENDIX A EXCERPTS FROM THE CONTRACT BETWEEN SUFFOLK COUNTY AND FUTURE RESOURCES ASSOCIATES .3 P l D D D D D
APPENDIX A O EXCERPTS FROM CONTRACT BETWEEEN r
~~ SUFFOLK COUNTY and FUTURE RESOURCES ASSOCIATES INC.
- 3. SCOPE OF WORF.
3 A meaningful assessment of consequences of potential accidents at the Shoreham nuclear power plant requires: (1) assessment of the probability of severe accidests for the facility; (2) determination of potential scenarios,
'D time seguences, and magnitudes of releases for the associated accident categories of the plant; (3) analysis of the characteristics of accident generated radioactive clouds and the local meteorological impacts resulting from their transport beyond the facility boundaries; (4) assessment of 0 public health consequences resulting from the accident-generated cloud transport depending upon the level of protective actions taken by the affected population. The program outlined below identifies a proposed f
approach to assessing the potential consequences in accordance with the "Q above outlined requirements. A detailed outline of the Tasks to be l performed, as described below, is presented in Appendix A to this proposed scope of work. I D 3.1 Task 1. Review and Critique of Previous Studies of Probabilistic' l Accident Assessments for the Shoreham Facility I I The review will include a critical assessment of accident sequences O defined in the earlier, LILCO-supported SAI study of the probability of l accidents at Shoreham. The projected probabilities for accident sequences defined above will be reassessed. The phenomena associated with fission product releases within the reactor containment structure will be reviewed. O The factors used by SAI to estimate potential reductions of fission product quantities released from the molten core as a result of the effects of engineered safety features and other physical phenomena taking place within Realistic estimates (from an the containment structure will be assessed. O emergency planners point of view) will be made of fission product behavior within the containment structure prior to releases to the environment. Containment failure mechanism modeling will be examined and reevaluated and quantitative descriptions of fission p'roduct release characteristics to the environment will be established as part of the requisite in7ut parameters for Task 2. O A-1
O 32 Task 2. Raview and Critique of Previous Radioactive Plume Transport Modeling and Radiological Consequence Assessments O This review will include an analysis of earlier LILCD-supported PL&G studies within this topical area and a reevaluation of the critical parameters associated with the transport of radioactivity and consequent O public health effects. An assessment will be made of the results of plume transport modeling and resultant radioactivity concentration studies. Critical factors used in the studies for the evaluation of health effects from the radioactive transport will be assessed. Estimates of population 'O distribution for both current demographic features and projections of future distributions will be evaluated. The meteorological models used in the PL&G analysis will be reassessed and the potential applicability of the models to provide insight into peak public consequence conditions that influence the 'O Protective shape of probabilistic consequence curves will be analyzed. action procedures and models will be analyzed and Long Island's unique ' terrain related influence on public health effects assessed. 3.3 Task 3. Performance of Site Specific Consequence Analyses and Assessment of Protective Action Effectiveness 1 Independent calculations of public health consequences will be made lO with the CRAC-2 code to assess the validity of LILCO-supported analyses. Preliminary calculations will be made with input parameters similar to those used in the utility-supported studies in order to develop an understanding of dif ferences in results induced by differences in numerical methods used Following O in the proposed studies and those used in the earlier analyses. the preliminary review of utility-supplied results, site and emergency response plan-specific analyses will be made to determine the consequences Other of reactor accidents if only minimal protective actions are taken. calculations will be made to estimate the plan-specific effectiveness and support the development of potential protective actions proposed for utilization by plat iers developing the County's Radiological Emergency Response Plan. An assessment will be made of the ranges of uncertainty
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A-2 O
lO ' associated with the results based upon the uncertainties of selected critical variables that may strongly influence public health impacts of C) nuclear power plant accidents. 3.4 Task 4. Project Integration, Documentation, and Technology Transfer 13 Support will be provided to assure integration of the consequence analyses with the activities of PRC-Vorhees, the County's contractors responsible for developing its radiological emergency response plan. Local issues and concerns will be assessed and integrated into the consequence l) analysis through support provided to the Radiological Emergency Executive Steering Committee and Suffolk County officials. Support will also be provided to the Steering Committee and County officials in presentations l explaining the methodology and results of the Consequence Analysis studies. C) Internal integration of the program will be provided to assure that portions of the study conducted under the direction of Dr. R.J. Budnitz and those conducted specifically under Dr. F.C. Finlayson's direction are properly coordinated. Overall supervision of the project will be provided by I) Dr. F.C. Finlayson. Documentation for the Consequence Analyses will be provided in a final report. The report will contain a presentation of the significant data derived in the study in a format useful for planners. A i review and analysis of the data vill be presented in the report. Critical lC) issues associated with the results will be identified and ..... sed in terms of the results that have significant impact on the County's emergency response planning for public risk reduction from potential accidents at the Shoreham nuclear power plant. () . C) l l C) l A-3
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g-APPENDIX A. DETAILED DUTLINE OF PROPOSED SCOPE OF WORK Ta sk 1. Review and Critique of SAI's PRA Treatment of Accident Sequences, r)
" Fission Product In-Plant Release and Interaction, Containment Challenge Analysis, and Analysis of Fission Product Releases from the Plant.
1.1 Critical Review of Accident Sequence Definition 3 1.1.1 Selection of Accident Initiators o justification for those omitted o criteria fer inclusion o ef fect of these judgment s on the final results O 1.1.2 Event-Tree Definition o system description i o justification for level of detail o treatment of common-cause issue s o human factors issues in event-tree logic
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1.1.3 Fault-Tree Development o selection of which f aults requires fault-tree analysis 3 o level of detail o human factors issues, including operating procedures
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o special attention to control systems issues and support sy stems issues 4 1.2 Quantification of Accident Sequences i 1.2.1 System Failure Probabilitie s
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o component failure analysis and data o human f actor s is sue s dependent f ailure analysis o Event-Tree Quantification
- ) 1.2.2 1.2.3 Fault-Tree Quantification 1.2.4 Uncertainty Analysis o systems interactions
() o dependencies across systems 1.2.5 Critical Review of Conclusions o special treatment of possible major issue s o uncertainty of emergency planning development to
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the se re sult s A-4
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O. 1.3 Within-Containment Phenomena For each "important" accident sequence , O 1.3.1 De scription of Phenomena Leading to Core Degradation o quantitative de scription, including time-sequence issue s 9"*"* i"*
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O 1.3.2 Meltdown and Release of Fission Products 1.3.3 In-Vessel and In-Containment Fission Product Behavior
*ff*** f '"8 i "****d **f**Y f * * * "' ' '
O o time progre ssion o uncertainty analysis 1.3.4 Containment Challenge Issue s 1.4 Containment Challenge Analysis 0 1.4.1 Time-Sequ'ence Analy sis of Containment Challenge o overpre ssure o bottom containment penetration steam explosions O o etc. 1.4.2 Containment Failure Modeling o categorization of accident sequences into groups O o f*il".*"*1analysts uncertainty Y 'i' . ! 1.5 Quantification of Release s l 1.5.1 Quantitative De scription of Release s lO o time sequences
- o duration of release l
o thermal character of release o elevation or site of release l i o fission product characterization O 1.5.2 Quantification o En.phasis on fission product release fractions o time sequences o phy eical properties of the release o uncertainty analysis 'O l A-5
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O Ta sk 2. Review and Critique of Previous Radioactive Plume Transport Modeling and Radiological Consequence Assessments 2.1 Plume Transport Modeling and Radioactivity Concentration
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o plume modeling parameters o dose-radioisotope concentration relationships 2.2 Health Effects Analysis O o dose conversion relationships o early deaths and injuries o latent effects o long-term exposure mechanier s (out of scope??) (ingestion pathway EPZ to be defined??) O 2.3 Population Distribution o current census (1980) o projected population distribution (2000(?)) , 2.4 Meteorological Models
.(3 o long-term historical data from site o single year hourly data source model from site o regional meteorological models (CRACIT satellite stations) o data preprocessing routine impacts (CRAC-2 & CRACIT) o sampling frequency requirements and
() probability / confidence results 2.5 Protective Action Procedure s o nominal exposure relationships without protective actions o evacuation models and site characteristic s () o sheltering models for local residential / commercial structure s 2.6 Modeling of Island-Specific Terrain Influences on Consequence s i I 'O Performance of Site Specific Consequence Analyses and Asse ssment Ta sk 3. of Protective Action Ef fectiveness i 1 3.1 Asse ssment of Utility-Supported Analy se s I o CRAC-2 simulation of PL&G input parameters () o comparision of CRAC-2 and CRACIT results 3.2 Site Specific Analyce s l o asse ssment of nocinal results without protective actions o specialized protective action ef fectiveness asse ssments
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3.3 Uncertainty Asse ssment for Critical Variables o assessment of variations in results associated with O uncertainties in critical variables identified in Tasks 1
&2 Ta sk 4. Project Integration, Documentation, and Technology Transfer D 4.1 PRC Voorhee s/ Consequence Analysis Interface 4.2 Suffolk County and Radiological Emergency Executive Steering Consnittee Support o Presentations to Steering Committee / County of Methodology and Re sult s of Consequence Assessments I
! o review of technological features of consequence analysis and implications to planners o reviews of final results of consequence assessment study o Additional Support as Required
! 4.3 Internal Integration of Parts 1, 2, & 3 4.4 Preparation of Reports o data presentation, review & analysis o issue identification and assessment o presentation of critical results of consequence
! analysis affecting emergency planning potential for public risk reduction ) i l l l 5 . 1 ? D A-7 h
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'O lO APPENDIX B i
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1 9 40
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Oi 'o ) I
- Q I APPENDIX B i-
- O REVIEW OF WITHIN-CONTAINMENT PHENOMENA AND CONTAINMENT CHALLENGES i
l Reported in
- Q Probabilistic Risk Assessment, Shoreham Nuclear Power Plant Station i
lO This Review Was Performed by Dynatrek, Inc. Under Subcontract from Future Resources Associates i, lO July 1982 t !O } 1 i j 1
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O CONTENTS D Page
1.0 INTRODUCTION
1 1.1 Listing of Documents Reviewed During the Study 2 D 1.2 Brief Summary of SAI Computed Accident 4 Scenarios Considered in This Review 2.0 ISSUES CONSIDERED IN THIS REVIEW / STUDY 18 0 2.1 Meltdown Process with RPV 18 2.2 Vessel Melt-Through and Discharge of Molten 21 Debris into CRD Room O 2.3 Transport of Molten Debris Within CRD Room 30 2.4 Study of Countercurrent Flow Limitation 40 Within Downcomers 2.5 Numerical Evaluation of Debris Transport 60 [) in CRD Room 2.6 Brief Review of In-Vessel and Ex-Vessel 63 Steam Explosion Probability 2.7 Overall Conclusions and Recommendations 65 0 3.0 IMPACT OF RESULTS OF THIS STUDY ON SAI RESULTS 67 CONCERNING ACCIDENT CONSEQUENCES O 4.0 LIST OF OTHER REFERENCES PERUSED DURING THIS STUDY 69 APPENDICES I Listing of Computer Program HTRANS Developed for 70 [) This Study and Results II Listing of Computer Program CMELT Developed for 74 This Study and Results O O
'O -
1.0 INTRODUCTION
Future Resources Associates, Inc., of Berkeley, California, was contracted-by Suffolk County, New York, to carry.out Task 1 of the " Consequence Assessment for Suffolk County Radiological Emergency Response Plan" related to the Shoreham Nuclear Power Station. The Dynatrek, Inc., was, in turn, subcontracted by Future Resources Associates,-Inc., to review and study Sections 1.3 and 1.4 of Task 1, titled "Within-Containment Phenomena" and " Con-tainment Challenge Analysis," respectively. Ten man-days (80 man-hours) were allocated for this study, plus the additional 5 man-j- days for preparation of the report. This report represents the Dynatrek, Inc., contribution to , the above-mentioned Consequence Assessment.
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2 O 1.1 Listing of Documents Reviewed for This Study Probabilistic Risk Assessment, Shoreham Nuclear Power Station, Long Island Lighting Company, by Science Applications, Inc., San Jose, California.
- 1. Volume I: pp. 1-1 through 3-12 and 3-138 through 4-6.
- 2. Appendix C: Containment Response Analysis for Degraded Core Accidents, pp. C-1 through C-90.
- 3. Appendix H: Containment Event Trees, pp. H-1 through H-112,
- 4. " MARCH (Meltdown Accident Response Characteristics) Code Description and User's Manual," NUREG/CR-1711, October 3
1980.
- 5. " Interim Technical Assessment of the MARCH Code,"
J. B. Rivard, Project Leader, Sandia Report, NUREG/CR-D 2285, November 1981.
- 6. " Core-Meltdown Experimental Review," prepared for USNRC by Sandia National Laboratories, Nuclear Fuel Cycle D
Programs, NUREG-0205, SAND 74-0382, March 1977.
- 7. Reactor Safety Study: Section III, Appendix A, NUREG-75/014 (WASH-1400).
D
- 8. "Phenomenological Assessment of Hypothetical Severe Accident Conditions of the Limerick Generating Station,"
by R. E. Henry of Fauske Associates, Inc., in Volume II ] of Probabilistic Risk Assessment, Limerick Generating Station, Philadelphia Electric Company, Docket Nos. 50-352, 353, March 1981. O m ~)
t' 3 O
- 9. Memorandum: L. S. Tong, Chief Scientist, USNRC/RES, to !
l R. B. Minogue, Director, USNRC/RES, " Technical Review O i Meeting-on Steam Explosions," with enclosures, May 24, 1982, USNRC Public Document Room. O t ) !O i o i !O a j !.O !O O O O O L
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4 g 1.2 Brief Summary of SAI (Science Applications, Inc.) Computed Accident Scenarios Considered in This Review q For the Shoreham PRA (Probabilistic Risk Assessment) , five representative accident sequences have been developed for detailed deterministic containment response analysis. SAI has determined O that these sequences represent the range of core melt events described in the probabilistic event trees. It was outside the Dynatrek, Inc., scope of work to either examine or challenge the choice of these limiting scenarios. Instead, our main emphasis I is on the. review of the methodology and of the results of SAI studies pertaining to (a) in-vessel (core / damage and meltdown O progression), (b) transport of core debris from the reactor pressure vessel (RPV) to the control rod drive (CRD) room, (c) transport of core debris to containme'.lt drywell and wetwell, and
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, (d) the resulting challenge to the containment integrity for the
.five scenarios selected by SAI.
It should be noted that the Fission Products transport O
- within the RPV, within containment, and to the outside environ-ment was also outside the scope of Dynatrek's work assignment.
In this Section, the salient results of the SAI determinis-O tic analyses, performed with the MARCH computer code (Reference
- 4) are summarized to allow for more orderly, subsequent descrip-i tion of the impact of Dynatrek's assessment.
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5 O 1.2.1 Class I Scenario: Loss-of-Coolant Makeup to Reactor a) Main Assumptions: o Some transient causes a reactor scram from 100% power Feedwater to RPV is unavailable O' - Emergency Core Coolant is unavailable Main Steam Isolation Valves (MS1V) are closed The specified operator action to depressurize the
/3 RPV is not taken. Hence, the reactor pressure reaches the pressure relief valves set point and the steam passing through these valves (SRVs) is O dumped to the containment drywell.
b) SAI Computed Events: -O Time Events 0-0.68 hrs. Core heatup at constant pressure, core gradually uncovered, with attendant clad oxidation. 0.68-1.0 hrs. Partial core meltdown. Core slump-ing into the RPV bottom head, at t=56 min. when more than 70% of core is melted and 22% of cladding has undergone chemical reaction (oxi-O di2*d) - Increased steam generation due to boiloff of lower plenum (LP) water; core debris quenched (may have become partially or totally solid). ~O 115 min. Reheat of core debris and some heatup of RPV bottom head. The latter fails due to high fluid pressure, combined with the elevated metal temperature, rather than due
- (3 to melt-through.
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6 ) Time Events Temperature of the molten debris y exiting the RPV is low. May be quenched in the drywell to below the 2500 F criterion for initiation of concrete interaction.
" Dead water" associated with recircu-p lation piping and other RPV fluid (steam and H ) are discharged into 2
containment During boiloff of drywell water, the core debris is quenched from 2600 F p to 732 F. Subsequently, core debris interacts with the concrete, caus:.ng genera-tion of large amounts of noncon-densible gas (mainly CO 2 ) and p further oxidation of the debris metal (H 2 evolution). 595 min. Containment pressure failure (120 psia) reached. p Concrete penetration continucs until metal solidifies, at which time gas generation is greatly reduced.
+ foNhyNMGNT PAEnuae p /10'
(/RA) 79 9 50 1
- 20. a
. ; - 7/ME 000 300 600 (Muvuras )
7 D 1.2.2 Class II Scenario: Loss-of-Containment Heat Removal a) Assumptions: D Some transient causes a reactor scram from 100% power Feedwater to RPV is unavailable D MSIVs are closed Loss-of-containment heat removal due to failure of the power conversion system and of the residual D heat removal system RPV is at the SRV set point pressure with steam being discharged into containment wetwell. O b) SAI Computed Consequences: Time Events D RPV water 1-evel decreasing. The reactor core isolation cooling system (RCIC) (using water from the condensate storage tank at 100 F) is initiated when low level mark is D reached in RPV. 88 min. Containment suppression pool tempera-ture rises. When T 120 F, the AutomaticDepressurE32b=ionSystem (ADS) is activated and RCIC injec-D tion is terminated. It should be noted that the reactor manufacturer specifications call for gradual rather than sudden depressurization as pool water heats up, thus pro-longing RCIC injection. D RPV depressurizes from 1146 psia to 50 psia in 3 minutes. D D
8 Q Time Events Water in the RPV flashes and the core is totally uncovered for less $) than 10 minutes because the Low Pressure Core Spray System (LPCS) is activated upon the drop of the RPV pressure. Adequate core cooling by LPCS. O water level increases and covers the core. The RPV is maintained at low pressure (~150 psia) by relieving steam through SRVs to wetwell pool. Containment pressure increases. O 25.9 hrs. Containment failure pressure (120 psia) reached. ECCS assumed lost due to loss suction head caused by containment g pressure drop to 1 atm. Water level in RPV begins to decrease due to loss of ECCS, and the core uncovers. Low level of decay heat. 27.8 hrs.
) Core meltdown initiated.
28.5 hrs. 75% of core molten when core slumps into RPV bottom head. 21% of Z oxidized. 30.0 hrs. Dryout of water in RPV bottom head.
)
RPV head attack initiated when debris reaches 1840 F. 36.5 hrs. The RPV bottom head fails when 27-53% of the bottom head is molten. () Core debris reaches 3600 F. Debris-coolant interaction on drywell floor and in wetwell pro-duces a 30 psi pressure spike, j) despite failed containment. 36.7 hrs. Core debris totally quenched to saturation temperature. O .O
9 O Time Events 43.75 hrs. Concrete-debris interaction starts O when debris reaches 2500 F. 46.17 hrs. Vertical penetration rate into concrete increases significantly when oxide and metal layers exchange position. O Decomposition gases blow through melt and react with metal. Heat of reaction increases metal layer heat source by a factor of 60 over the decay heat.
;O 47.75 hrs. Metal layer frozen, vertical pene-tration of concrete terminated. ~O
{pNTHNMENT tWERWRE (FCIA) to ./zo - .O 40 O , l 20-l .o. 25 9 36 5 47.gr NOWRs l 1 u auraver FMLs \O gPV l FAnwat i i l0 l
10 g 1.2.3 Class III Scenario: Large Break LOCA a) Assumptions: 'O - Double-ended (guillotine) break of the main coolant recirculation pipe, upstream of pump All ECCS lost n-* - Pressure suppression pool cooling is unavailable Containment sprays not operational b) SAI Computed Consequences: O Time Events 1 minute RPV pressure decreased to the con-O tainment pressure. Core is uncovered before the fuel rods can heat up ' significantly. 18 minutes The lowest core layer does not remain solid due to lack of cooling. m Core starts to slump into RPV ^# bottom head when 51% molten. 0.35 hrs. Bottom head water rapidly evaporates (containment pressure reaches 80 psia) and RPV bottom head attack is (3 initiated.
- High energy gases produced during core slump cause very high drywell temperature (3000 F).
"() Very high metal-water reaction rates computed for the intact core region i after slumping due to assumption of no' flow blockage. Steam starvation could increase the i O meltdown time period of the (ini-tially) intact core region by about a factor of 10. Rapid heatup rate of RPV bottom head. Debris / wall interface climbs O from 370*r to 2600 F in 2 minutes. O 1
11 O Time Events 48 minutes RPV bottom head fails when 52.5% of O wall thickness penetrated by the melt. Vessel inventory plus 32,800 lbs of water (from recore piping) dis-charged into drywell. O Hot core materials interact with 126,100 lbs of water on drywell floor and containment pressure increases to 95 psia. () Fragmented core debris quenches to 1175 F on drywell floor. Water limited interaction. 53 minutes Drywell water evaporated. Total metal-water reaction of 37% produces 13 2050 lbs of H However, less than 2%producedd$r.ing core-water inter-action outside RPV. Subsequent steam generation rate very low (10-150 lb/ min), causing-
- (3 containment pressure to drop.
274 minutes Core debris on drywell floor heats up to 25 F, starting interaction with concrete. C) 324 minutes Strong interaction with concrete, due to metal vs. oxide layer re-versal, causing large evolution of noncondensible gas. 354 minutes Containment pressure fails (120 I) psia). 474 minutes Melt propagated 11 inches into concrete then stopped. Event terminated. I
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- O 1.2.4 Class IV Scenario: ATWS
.O a) Assumptions:
Some transient, at 100% reactor power MSIVs assumed to close O Feedwater to RPV is not available . One recirculation pump trips. Due to 50% reduc-tion in core flow, neutronics feedback reduces
- O power to 30%, as computed by the manufacturer.
Failure of control rods to insert Failure of the liquid poison injection system O
13 J b) SAI Computed Consequences: ~) Time Events HPCI turns on at low RPV water level. Throttles back when high level regained. RPV at high pres-sure relieving steam to wetwell pool 7 via SRVs. Capacity of the Pesidual Heat Removal (RHR) system is insufficient to remove the additional heat load. J 32.4 minutes Containment fails upon reaching 120 psia. HPCI pumps cavitate dtte to low Pcont* 3 Core starts to uncover; power starts to decrease due to void feedback. Core uncovers fast, hence, rods do not heat up significantly, causing 9 low amount of clad oxidation. Water level drops by 14 ft/ min. after ECCS terminated. Water level is at 2 ft above bottom G support plate when core heatup occurs. 46 minutes Core meltdown starts. Steam genera-
- tion low since core nearly totally uncovered (water level less than 0.5 9 ft).
7% of clad oxidized and 55% of core melted when core starts to slump into bottom head. Noncoherent slumping since no water present. 68 minutes Total core collapse. Prior to that, rapid avergy transfer due to partial slumpi.ng, vaporizes all water in the RPV lower plenum. 47% of Z O*i" r dize6 since no flow blockage assumed. ..) 7 w/
14
.O Time ' Events
. Quenched core debris and-support-
.O structure.in lower plenum come to thermal equilibrium at about 3100*F.
4 73 minutes Core debris starts to react with
-- RPV bottom head.
C) 105 minutes - Bottom head failure without melt-through caused by 1075 psi pressure differential and elevated wall' temperature. Following RPV failure, containment I)- pressure increases in spite of-ruptured containment. , Energy usurces: 51,300 lbs of saturated water in recirculation piping plus 34,000 lbs of steam at .O elevated pressure in RPV. vapor space. 110 minutes Further debris-coolant interaction j outside vessel creates additional energy source causing-second pres- .O - sure spike of about 120 psia. 118 minutes Drywell floor dryout. Core debris quenched during dryout from 2600*F to 434*F. I) 534 minutes Total period of about 400 minutes. t during which no significant mass and energy generation from debris-coolant or debris-concrete inter-1 action, except for vaporization of wetwell water due to fallen debris. i Concrete penetration starts when i oxide layer becomes molten, followed by' rapid degassing. The latter becomes'the main driving force for fission product leakage to atmosphere. I) ; i 4 j. lO _ , _ _ , . _ . , . - ..,_-_.,___-,___,.s ,.%,_ _ ,_ _ _ . , _ _ _ ,_ . , -_ _ _ . _ . ,, . ___,,_ _.,..,_,-.,7__. ,_,_,_,._-._-_,._,,._,........-cm.,... . - _ . . , _ , . _ _ , _ , - . _ _ _ _ , , , . , , - . . . - , , - .
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d- ; . . ; ; 100 Z00 .I00 4 00 N/Nyrrs ,0 1.2.5 Class V: Main Steam Line Break Outside Primary Containment: Containment Bypass a) Assumptions: Main steam line pipe break (0.678 ft 2 break flow area due to flow restrictor) outside primary con- , tainment when reactor at full power. l Isolation valves (MSIVs) fail to close Coolant makeup system failure Extended exposure to high temperature steam within secondary containment causes ECCS equipment failure Secondary containment blowout panels open when 3 psia reached. O
16 O b) SAI computed Consequences: O Time Events Amount of clad oxidation as per Class I. C) Primary system boiloff bypasses containment from beginning of transient. 5 minutes RPV pressure dropped from 1026 psig to 273 psig. O 12 minutes Core starts to melt 10 minutes after uncovery. 16 minutes Core slump init!.ated when 60% of core is molten. O Core debris is held up in the core support structures and no vaporiza-tion predicted after initial energy exchange until after the second support grid reaches its failure 3 temperature. Hence, metal / water reaction in the core region is limited. 50 minutes Bottom head attack begins. 3 209 minutes Bottom head fails. Core debris is at about 3780 F (2082 C). Debris drops onto dry drywell floor. Hence, no debris-coolant interaction in drywell. 3 Core debris interacts with structure in the reactor pedestal area before interaction with concrete. I 300 minutes Debris-concrete interaction. Degassing releases energy into O' containment which is initially ' absorbed by wall and other internal structures since these were cold. 3
17
-O Time Events 400 minutes Rapid penetration of concrete g starts. Energy relegsed into dry-well is about 2.8x10 Btu / min of which 80% is absorbed by contain-ment walls.
Containment drivell fluid (gas) O temperature reaches a peak of
.. 370 F.
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18 O 2.0 ISSUES CONSIDERED BY DYNATREK, INC. 2.1 Meltdown Process Within RPV O It is our impression that SAI engineers have used competently and conscientiously, the only tool available at the time; i.e., the MARCH code, designed to address, in a bi lhly simplified
-O manner, the extremely complex physical processes involved in the meltdown scenario.
The extensive simplifications in the code were introduced for O two reasons: (1) to produce a tool useful for probabilistic risk assessment which requires many computer runs for exploration consequences of various bounding assumptions. Hence, computa-
-o tional economy must have been one of the principal goals; (2) to produce a tool amenable to accepting some of the major uncer-tainties as input assumptions that could be changed from run to O run. These stem from the relatively poor knowledge of various thermohydraulic processes that involve melt propagation and the attendant heat and mass transfer in complex geometries. The -O pertinent empirical base lags far behind the existing empirical base collected in the course of reactor safety research for situations that do not involve a degraded core. However, the word
.O j " uncertainty" is also used here to imply simplifications one needs to make to intentionally bypass the detailed calculations that I would cause the code to become long running. lO Sandia engineers have performed a thorough review of the MARCH code (see item 5 in Section 1.1) and found many items that l could be improved upon, most of which were known to the Battelle O As Columbus Laboratories staff that developed liARCH for USNRC. l l lO i s
19 O i a result of these and other reviews, the USNRC has sponsored the development of a significantly improved and extended code, now Q, named MELCOR, the development of yet another code at the Idaho National Engineering Laboratory (INEL) named SCDAP, and procure-ment of more pertinent empirical base at INEL (in the PBF test facility) and elsewhere. When reviewing the major assumptions in the tiARCH code, one gets the impression that the code was primarily oriented towards pressurized water reactors (PWRs) and their containments that do not employ pressure suppression systems, such as in Shoreham. The RPV structure, its flow patterns and cooling modes, and, u especially, the structural components within the RPV bottom head, are quite dissimilar from PWRs. None of these have entered in any significant way to the MARCH code application used for the Shore-ham PRA. In the course of the SASA (Severe Accident Sequence Analysis) ! calculations being performed by the Oak Ridge National Laboratory .O (ORNL) for the USNRC as applied to BWRs, the ORNL engineers have also come to the conclusion that MARCH has many deficiencies. Out of 21 listed inadequacies, we shall extract only a few:
- Modeling of core heat transfer to upper and lower BWR
! vessel internals
- Modeling of core collapse
,s %J
- Failure of bottom head via CRD tube penetration not considered
- Suppression pool and wetwell/drywell interaction
)
O
20 ) Rod-to-rod radiation heat transfer not included Vessel water level calculation does not include vari- ) able flow areas Fuel pin melt / slump / freeze phenomena are not mechan-istically modeled. ) The work on removal of the last item mentioned above was subcontracted to the Renssalaer Polytechnic Institute (Professor R. T. Lahey, Jr. , as Principal Investigator) . * ) Some of the listed inaC2quacies have already been removed by ORNL and/or Battelle engineers. Others may take years to replace. Dynatrek, Inc., staff could not undertake any serious ) effort in the 10 day-assignment to address consequences of these code weaknesses and have, therefore, decided to accept the SAI computed in-vessel events at their face values. We would have ) felt more at ease, however, if SAI had also considered the inflow of the CRD (control rod drive) cooling water at 185 gpm, since it could significantly delay the in-vessel meltdown and debris . ) collapse onto the RPV bottom head. ) ) )
21
.O 2.2 Vessel Melt-Through and Discharge of Molten Debris into CRD Room 'O 2.2.1 Properties of the Molten Debris, Pertinent to Shoreham UO : 456.7 lbs/ assembly x 560 assemblies = 255,752 lbs 2
O Z r: 224.3 lbs/ assembly x 560 assemblies = 125,608 lbs Steel: If all internals were to melt: ~500,000 lbs Total weight of "corium" 2400,000 kg O Weight % UO2 = 29 Z " l4 r SS = 57
.O Corium properties:
de*t/$ 0 = /p f0 Zf a /p f( t G /4 , l-( f p.g7 , 7.g y (.j,,) 2 fpc $& L = fD29 /2f + p M *5[l f 0 r7 v 47o) u 6/3 9 'c
- O w ./. w 4 fa <, a )
I g f0'l f r 0 037 t 0 7/ 0'3 ) sc 0 22 f o, = 22 /, .,g { O Heat of fusion, /)_ = Z 7 / E v Y( O O i O
22 0 2.2.2 Melt-Through Mechanism We do not believe that the RPV bottom head would fail in 9 structural either of the two ways described in the Shoreham PRA: failure of the vessel wall due to elevated pressure and temperature or the wall melt-through when the RPV pressure is low. O Instead, we agree with the scenario described by R. Henry of Fauske Associates, Inc. (see item 8 in Section 1.1) wherein the relatively thin metal that seals the CRD tube, or the CRD tube ' ,0 itself, is much more likely to fail first upon contact with the melt. The melt is very unlikely to reach the bottom head in a
- O coherent fashion (gross slumping). It is more plausible to con-sider downward streaming of melt around the vessel axis. That melt will not attack the CRD tube as long as there is some water left in the lower plenum. If the amount of steam generation caused by quenching of melt is insufficient to stop further melt-ing (a likely case), increased amount of debris will accumulate on the bottom portion of the RPV and it would eventually remelt the fraction of debris that was frozen by the liquid--which by now has evaporated. This whole process could be delayed significantly if the CRD cooling water were continuously supplied.
Eventually, one or more CRD tubes would fail and the debris e discharge into the CRD room would commence. O o O w
23
.O 2.2.3 Melt Discharge Rate at High RPV Pressure Since it is anticipated that the total discharge of the melt
- O would last a short time and the melt volume is small compared to the total system volume under pressure, it is reasonable to assume that the fluid pressure inside the vessel will not change until all O
melt has discharged. As the hot debris flows through the initial opening (of assumed equivalent radius rg), the RPV wall would ablate, thus, enlarging the break size. Let: h = heat transfer coefficient between the melt stream and the vessel wall
- O T, = debris temperature (-2000 C)
'f s,m = m lten steel temperature (-1500*C) p = vessel wall density s
T = vessel steel (initial) temperature (~300'C) s pRPV = 7.0 MPa pCON = 0.2 MPa
.O The rate of ablation / melting of the vessel wall could be l
l- found, approximately, from l Z Og (F., -7;)a, prs p = zul tt-r ,) ss A fr - r-,) 0
~
#* 4 E h M - 7; ) A s J d) l O
tO
24 O R. Henry utilized the Reynolds analogy to determine the heat transfer coefficient O 4 it's 9. x-z O where f = friction factor ~0.005 Nm *
~{l ,
= 40 fe . - (z) o k c= S2/rN -
-y lg ~--- E ' 7/ r /* 2,/
O and gn - s u . d (z - - ir 0) W 7*Yf~'/"f:79 f/m- E90) f 2 767t!]
= o oso ,
D (9 A [j) = /2, + D V10 7 - l - (4) Let n * "" CRD ' less simultaneously. Then, the total break area j, N) = n,,, h ' p nn-d)' -
/r)
D D
25 3 Let B = 0.0307 (using Henry's nomenclature), then, since g = 40 m/sec (from equation 2) , volumetric flow of melt- through the total break beccaes Q N) - 40 m,,7 A n + )*. s - d) ') where CD = discharge coefficient ~0.6. Mass of metal / debris discharged 3 g k A) = f_ lQIO M 3
=l'"!vo'G^%,61,*/+A,3/'1ft') .
/;)
g Let n CRD " ' D"
- B = 0.0307 J,
rg
=fx0.1=0.05m
}f,' //):: 3 Zdrk' / fz Sr/* / / 75fa/E-l + E /4/t v/od# )
O t(secs) g(kg) b RK(* } 4 g 1 1.41 x 10 0.1023 4 2 4.4 x 10 0.1949 4 3 9.65 x 10 0.317 4 1.77 x 10 0.469 9 5
'S 2.92 x 10 0.65 O
26 O Not all of the available metal within the RPV will melt and flow out. Assuming that about 50% will flow out (i.e., ~200,000 lJ kg), it could discharge in'less than 5 seconds. 2.2.4 Gravity Driven Discharge of Melt To simplify the matter, we will assume t'lat all of the debris that could flow out has been accumulated in the RPV bottom head just prior to failure of n u es. CRD The height of the molten debris within the hemispherical bottom head, as shown by Henry, can be found from /dn's v.,/mf Q N
~" D ..
{p) o . where R sin 0 = 4 R cos 0 = R - h 2 r = 2Rh - h , where R = radius of the RPV
}/ = 7/~ ff M ' N . .
fy} i ! The rate of change of that volume will equal the volumetric ( outflow Qa>= dj =-rn 6x-") $" . ri) ?.
27
'O i
velocity of discharge through the break b M H '
=Y2f . . . . fi,)
Therefore, py " h:go t Thff)isf2f//0) - (/z) As before, i O efs_ d IL .T, ) E ~ J][9, lE -7;) +?4 7 O where now
/= ~ Y'EN .
f,g lO and the Reynolds analogy was again utilized, together with equation (11), for the jet velocity. 'O 4i Ka N ' 9~ #'~ "~ ) N ... (is) 4Es hi - t > < r .7
=N5 ..
/M.e) 1 /
IO l // //) = A f k YN/) **'t ... f/C) i a l l O 1 i i l
28 O Equating (10) and (11) and utilizing (15) e O t NIZA-N) = - Q % ,, )/zyW A, + k /WN ... (h) O g
$2A-N)h#= - 4 ^%, IT7 fAa + Kll tow l O
This equation can be solved numerically for H(t) , utilizing forward finite differencing and small at O
$
- N, '
fu-a )m: l l4tkl ,,
&l ... (h)
- o where, using trapezoidal (simplest) numerical' integration 3 a ,L t = mat 6
l5N
- h & f &) + f$f S) + k, f-k )
O
- O and /fi = initial height of melt in RPV.
O
- O
29 9 Another way would be to solve for r using (14a) by forward finite differences J
& =k., h 4 l 'k lQ,, '-
kf) D Substitute on the RilS of (16a)
, /L bx- e.) "~ f-' =- a n,, m)A., w & E )'
f=k, fh ,,,, A d f k' f , ) . fq) l' E - M ., ) M ., D With r and 11 known from (18) and (19), respectively, O ppy is found from (12); providing fine enough time increments are used. D D 0 l ,0 l I
,w ~J Lg
30 0 2.3 Debris Transport Within CRD Room 2.3.1 Debris Outflow from CRD Room The floor of the CRD room (pedestal room?) is provided with four downcomers that connect it to the wetwell pool below. The remaining 84 downcomers (six of them capped) are located on the O drywell floor area outside the pedestal. Within the CRD room, the upper edge (entrance) of the down-comers is less than 3 inches above the floor. In addition, the O CRD room is provided with a 2-ft wide x 6-ft high manway that provides a passage to the drywell region outside the pedestal. That manway is blocked over its lower 2 ft by a 2 -inch-thick steel O
" sill" that, presumably, would hinder transport of any fluid across the manway, unless airborne.
The CRD room contains drive housings for 137 control rods, ~O various pipes, and massive support beams, all fairly tightiy clustered below the reactor bottom. Any jet (be it pressure or gravity fed) of molten metal would be first intercepted by this 'O structure. It is unlikely that the structure would melt before all the melt within the RPV bottom head has been discharged. While some dispersal of the melt onto the drywell floor outside the O pedestal area may take place (through upper support structure), it is judged that this will be a very small fraction of the amount discharged from the RPV, in the case gravity discharge, and a somewhat larger fraction (still less than 5-10%) in the case of pressurized discharge. O O
l 31 O As the melt discharged from the RPV accumulates on the CRD J room floor, once its level exceeds the height of the downcomer top flange (about 3 inches = 0.0889 m), the molten debris will start pouring into the four downcomers. ,
, A simple " weir" type relationship could be used to evaluate J
the volumetric flow rate per unit length of downcomer circumference A NA) { b) =o 1/zyy dy = lGh40Q . (2,) where AH (t) =HCRD(t) -H TD IICRD(t) = Molten debris level height in the CRD room IITD(t) = Height of top edge of downcomer above floor in CRD room. It is shown as a function of time because it could eventually be removed by melting. , That relationship is valid only if the flow of melt through a the downcomer does not encounter unusual resistance as formed, for example, by the countercurrent flow limitation (CCFL) which will be examined in Section 2.3.3. Consequently, barring any further obstacles to downflow of melt, the total volumetric flux of melt leaving the CRD room through four downcomers becomes
/)= f
- A r 2 T Ag 4 //
as /, vi K Eg=/{/=GJo/M h b)= 22 /L &f)~ 8) . (zz)
32 ) In a similar fashion, outflow of melt from CRD room through the manway becomes 3 h) = l Naa ~ $igi E - (23) where H gygg is the manway sill height which can also disappear if the sill plate should melt when in contact with hot debris. D With b = 2 ft = 0.6096 m k A)= H % IO - bj,))] ... g,y Finally, the height of molten debris in the CRD room is 3 obtained from the following mass balance: A,; d,N* A?,/0 - Q,M- Q/0 e
=
(2d where A CRD = net floor area in de downcomer room. ] From Shoreham PRA specifications, the gross CRD room (cavity) 2 floor area = 261.1 ft , 3 2 2 7r 2 The area of four downcomers = 4 ): 4
= 12.57 ft A
CRD " ud
33 O Equation (24) will be solved numerically for HCRD ( t) for the cases of pressurized discharge and the gravity discharge. O To summarize, the source and sink terms: Pressurized DischarrJe: O ppy = 75.4 n CRD (0. 05 + 0. 0307./ )1 O as long as Mm(t) < 200,000 kg where, by equation (7) s O Mm( f) = 6.484x 10 .n CRD e(2.5x10-3 + 1.53sx10-3 t
+ 3.1416x10 ~4t2)
Q Thereafter, Q RPV
= 0.
Gravity Discharge: O &y N) = d' 3 S'%, A *5 i
/2 = G DT + /< EU O
and H is found from numerical integration of 9 fffN-//) =-2/N M,,'k g where 5.54 = 2R = 18'-2" RPV diameter and the initial melt height O in RPV, H g, is found by trial and error from equation (9) : 2 27 & '
= Nr ,_ 74 vrt
.O N, c= / JTS , 0ppy 'O ub.,sy=0 O
.O 0,t=226zh,,0.)~Q4) 'O
&l,,, = IT[Na, A)- Niiu 0)}i.c Bef re proceeding with examination of downcomer countercurrent O
flow limitations, it is necessary to determine the time behavior of HTD(t) and Hgygg(t). O 2.3.2 Melting of Steel Barriers Two cases will be addressed: The first involving an analytic O
" *" " ' " ""** ** * ' ** " '"' "'"""" 9' " '
steel; the second based on Hesson's model (see NUREG-0205, p. 5-30). O a) Analytic Solution for Semi-Infinite Steel Region o .' . . .
.. ' '. / /, "
O -Q1,;// _ Consider a very thick steel slab with one face being kept in intimate contact with the liquid debris. Let Tg(x,t) denote the local temperature of steel T,(x,t)'= local temperature of molten debris T = melting point of steel O m,S
- O
35 D Governing equations: D [27) H c.V' N ... D Di E = xDx' _ D ' r- ... (u) D UX- l)# & m X~ -4 (2 2) C lY, f) = Tn, af u-o, />o . tar) D N * {
- T~,s n A x = $ lI] .
(2 9) X (t) = position of the steel melt boundary D
"Zd ff)-$~p, x=' xcX e . .
/10)
D where k = thermal conductivity s = heat of fusion A s = steel density p
.2 = thermal dif fusiv!!y L lx. n = T., + 4 ef,,;, .
M satisfies (26) and (28) D [lx.f)= {, +E.ey'c,,*, -- (32) satisfies (25) and (27) O O
36 O Upon substitution into (30) O E, fs z w _ LA j .$. u.1
*/g 27 gy
0 . hu,6 Yru_t s 'O This will be satisfied, for all t, if fk = If ... (14) g where 8 is a constant that needs to be determined.(34) into (33): g%"' f E85 s
- 4. A 5 = zfp, Vx_
??hXs i YirM;/- 2 Vi
'O From (29) and (31) O Z , - r_ ,, A= - bi)
.&rff
- O From (29) and (32)
^~'
B= O Y'hh) , L g Then, since R= , substituting (36) and (37) into (33) O
.-3 37 v
.O 4
~
I=,s ~5,
$r y_
5 = h ds $ . . . /g) x, aps r.-r- L 4(sirf) p (r-.- r. D This equation is solved for 8 by trial and error. Then, equation Q, (34) can be used to approximate a time when a thickness T I"XI s will be melted. Hence, time to melt i i b by - h1 g 2 1 .O 3 l Let og : 8000(kg/m ) j c = 500(J/kg C) 1 m
.O A s
= 2.76 x 105(J/kg)
{ km : k, = 46(W/m C) l [ gm :Rs : 1.164 x 10 -5 g,2/sec) , T mo : 2200*C T = 1500 C ms D T so = 10 0 *C
\ -A' A' L - 2.o _4 = /. Att ..
(4) afe$
-enft$
By trial and error, 8: 0.275 g 4 i
,, . . - _ - - --e,, , . - . - - . , , - - , - - - - , , , , . , - , - - ,,.-sy
38 O l [,(inch) 7,(m) At, (see) At, (min) O 3/8 0.009525 25.8 0.43 1.0 0.0254 183.0 3.0 2.0 0.0508 733.0 12.2 3.0 0.0762 1649.0 27.5 ) b) The Second Model, Lumped Parameter Concept g According to Hesson's experiments (see NUREG-0205, p. 5-30), the horizontal heat flux from molten U0 ,2 to steel is about 2 30.0 cal /[cm - sec) = 1. 26 x 10 6 gj,2 . Then the lumped parameter , heat balance yields yn
.( Q E tt ,-7;) q /\ L = / uwda . . .
n) 'O With c : 500(J/kg 'C) p O Tms : 1500*C T so : 100*C 5 A s : 2.76 x 10 (J/kg)
- O 3
p : 8000 (kg/m ) s At : 6188 8, (sec)~ ,0 l^ iO .O
39 J3 [y (inch) J',(m) t (sec) t (mind O 3/8 58.9 1 1 157.18 2.6 2 314.3 5.24 () 3 471.5 7.9 This model gives, as expected, significantly shorter melt g times for steel thicknesses greater than 1 inch. While the above melt times may be appropriate for addressing the manway sill molt, the downcomer top flanges woula require g consideration of the downward-directed heat flux which Hesson found to be smaller by about a factor of 2.5. Consequently, a 1 -inch-thick flange, if covered by molten g debris, would melt in approximately 4 x 2.5 = 10 minutes. The 3 inch stub of 3/8-inch-thick downcomer pipe protruding above the flange would be removed in less than 1 minute by ablation, as soon .O as the melt starts to pour into the downcomers. During our visit to the Shoreham plant, we have noticed that each downcomer top carries a (missile shield?) steel cover plate .g about 1 inch thick, supported by three or four columns (about 1 inch in diameter). While these steel columns would melt quickly, the cover r) plates could either fall down and temporarily block the downcomer, or because the molten debris is heavier than steel, the covers could be pushed away by the jet or by debris before they are also o melted. O
40 O 2.4 Study of Countercurrent Flow Limitations Within Downcomers 2.4.1 scenario O As the molten debris pours over the downcomer entrance, it is likely to break up into chunks or globules. Initially, these gl bules encounter only the negligible resistance offered by O more/less stagnant steam. They would reach a velocity N"
/
.O by the time they hit water within the downcomer. g If E y + h = downcomer length and h is the submerged depth of the downcomer (= 8 ft per Shoreham PRA), then, for the total downcomer length of about 56 ft, .O b " #~"'~
/4, s /r Mc g As soon as the melt hits the water, it will break up into much smaller particles.
The final size of these particles, if they had many feet g (>>8) of water to fall through, could be obtained from the force balance on particles and the Weber criterion: db Th = ffa .()( f TA . l'4A) 'O ad We =f ll ?^~' c. (4.s) g where r,= radius of the representative particley l l 'O .
41 O C D = drag coefficient for spherical particles pg = density of water V = relative velocity between particles and water r pm = Particle density O a,= particle surface tension W e
= Weber number O
Combining (42) and (43) J r f~ k ( f $,.~$ f - (44) back substitution into (43) gives O 3 t. (f .),) OVA O With o m = 441 dynes /cm = 0.441 N/m 3 p,= 8.6 x 10 3 pg = 950 W e" INeI CRIT g CD: 0.5 l (8W
, )I : 2.8
{3 C 9 WhI: 1.5 j 32 ( 3CD) ( O V r
= 1.22 m/sec, r,= 3.714 x 10- 3, O
42 O . Due to the fact that the length of travel through water is much shorter than needed to obtain the terminal size, we assume O that the effective r,is about 50% larger; i.e., ! r,e 0.005571 m .O (particle diameter of 11.5 mm) As these particles traverse the water space, the heat trans-fer between particles and water will produce steam that would flow O upwards against the descenting particles in the-steam space above the water surface. That steam flow, under certain circumstances, may be large enough to generate the drag on particles that equals or exceeds the particle weight, thus preventing the particles from proceeding downwards or expelling them out through the downcomer entrance. That situation is called " flooding." O Presence of metal droplets within the downcomer liquid , region and their deceleration to terminal velocity, if attained. will cause depression of that liquid region.
~ The extent of that C
liquid region needs to be calculated since it affects the amount of vapor produced. VEtoctry of /tttr nnow l - supearicat O ^ 'I i j ficans I. -O gy v w g $", E "; " O L n To.?.Y i f * *) **
. u, . pe <nu e r.u arr
* '. E Txu wsru
, , V I g A
l g 43 I Static force balance on column "h" g /4t [orce due to deceleration of particles, per unit area, = { Fy = j2 PmI 1
~
2 I F2 = gravity force of metal / water mixture 3 = phg F3 " P Lb9 Force balance: Fy+F2=F3 Metal volume in h is 2*A p Liquid volume in h is Ap h - j2^p O Hence f = f, /f f/~ ) O =kfff,-f)b-g where we have ignored the fact that some vapor will also be present in h. U2 is the mean particle velocity in h. Therefore, j L. lu - 4. ) + //_-L My A + g <7 = g L 7 i 6 3 r, L - t h - s )2yj d= . (n) g +(L-4)k v. Time to traverse h, At = . . . (47) 3 3
g 44 Particle velocity, U 2, could be found from the force balance A ln
- c ' & Yh$ _
- f +-
h,.'. ~ )l )f
# =-l a f k + P;'s -
m where r,will be fixed to r, 0.00557 m C D
- 0.5 3
p m = 8600 kg/m 3 pg = 950 kg/m 3 .-
' = - 5 7z Vf + P 73 = f 73 f/- 0 4t Wl) et dWz
= - f 73 0 '
(4 q} 0 /.s kl-l
]
Let /0.45 E a = 0.6526 Integration of (46) results in D
/ d k,-/
2d d h, 4/
- f 7 / f l' ]
when t = 0, U2=Uy
,., fg WQ' _ fYN I i
) = - /7. Jid /
/
/ d k, +i j (Ak-// i dl<,-/ ) !'### 0 d k,-J _ (y d % +/ Ak t/ }
i 3 m/ l
45 3 Hence O d k,-) = (dh, +l )y20) D Nz 0,l= g/ f/,_4 (W/) ,,, fro) ,3 i When t = 0, [ = dN
/
~/
, _ d W + / 4 O,-/ _
i 2 d t<, d4 itl dk,+/
~
dbro +/
/+ /$
g
- dN i ar ~d, k ,' 4, , a s d ni w .
O When t + = , S+0 k, , = /.73 , fw Heeruumr O nrbin (t)# O With Uy = 17 m/sec, a = 0.6526, aUy = 10.64
-11.39t O S(t) = 0.83 e
- // se f
/f 0f3 4 H,N)= M5 _ ,,. , a (s)
/- till e 3
0
46 O-t(secs) B(t) U2 (m/sec) z(m) 0.01 0.741 10.27 0.14 0.02 0.66 7.48 0.33 0.04 0.53 4.98 0.45 0.06 0.42
) 3.74 0.54 0.08 0.33 3.03 0.61 0.10 0.27 2.65 0.67 0.20 0.09 1.83 0.89 )
0.40 0.0087 1.55 1.23
-4 0.60 8.9x10 1.53 1.54 0.80 0 1.53 ) 1.85 1.00 0 1.53 2.15 where z = distance travelled through water (by numerical inte-
- () gration).
As can be seen, the particle would reach its terminal velocity of abcut 1.53 m/sec in about half a second, at which time it would .() have travelled a distance of about 1.35 meters. If none of the liquid were expelled from the downcomer (through bottom exit) , it would take the part icle about 1 second to traverse the initial
- (D liquid region within the downcomer.
To simplify the matter, let U 2 *U2 = terminal velocity of particles = 1.53 m/sec. () Since L = 8 ft = 2.44 m, equation (46) becomes s , T
- 2 5 tr - /3 S%2 ft
= Z J4 /~J~fffz >
-(F2) ff0 f f* ft
/ f f 2Fl and At = 0.65h - -
If3) I(3 -
C) 47 Equation (52) also' implies that if j2 # 5 85 (m/sec), the
.() liquid region h would disappear; i.e., metal droplets would not
~ create steam within, but outside the downcomer. That steam would vent into the wetwell vapor space rather than through the down-
.() comers because of lower resistance, thus eliminating flooding.
2.4.2 Flooding Criterion
- 7) Let j y denote the superficial velocity of upward flowing steam kV<
yl # i /v >$s
- O W , = ;nass rate of steam flow A = owncomer flow area D
O p ,= steam density Let j2 = superficial velocity of particles 0 47
,L . . .i i
N As where O d was given by equation (22) for all four downcomers. !O At the flooding point, the following relation applies K 1 i . i +
'C = /d +
4 - Ms) .v where, according to Wallis (p. 383) lO g ~_ 4 l lQ
48 e and v, is the same as V r given by equation (44) except that p y replaces pg: C # 2 rf '![<r.. 7 (L -r )]' - (rn D Hence, from (44), setting Kg =Q .. [N) 3 $; 7
- l -
lF') Next, we turn our attention to the evaluation of j y from con-sideration of heat transfer. 2.4.3 Particle / Water Heat Transfer Ignoring chemical reaction heat and decay heat generati'on due ) to very short times of interest, the transient heat conduction equation for the metal sphere, transferring heat by convection to the ambient at T sat, becomes
~ = R_ 'f p - , D & st 4 G. -
Nf) ) where a = radius of the sphere Initial condition C /h,0) = Co ) )
49 .O Boundary condition
.O s s,a . Z jD +4'fr-G)=0 -
fn) where r=a O h= (constant) convective heat transfer coefficient Let 7 - T, - T sat 'O V^T -T mo se . Ad h' = h/A, Then O = 4 *f ' Yd
- A .3 y
{t .p. / 's , o e-l.sv s - A - hY"h A Analytic solution for this problem can be found in Carslaw and O Jaeger, p. 238:
.a l
gg'y ads a ug;% (eg './)* y-t's, /)= ^ 4 ha ad]& A4 O - = ' 4 T*2"<'f"4'-'8 (si) where a n, n=1,2,..., are the roots of aa n t(aa n) + (ah'-1) = 0 O ... fg2) The droplet surface superheat, v(a,t) , can be found from (51) by setting r = a, and the instantaneous heat flux to water is O f A)=-W ; 4. [Sh O
50
- O Let
.o " "
y n (zl.'V)e~"~ E ... (a) 4,'[a k.'+ 44 74 <'- d 7 O then -a vfa,i)=l [ B A l*4) a. 2 /
/tr)
O and, since A .o b SA. _ .I-h' 8 A = 7 . -[ d 8 MAda,
^ Ms O **'
from (53), we obtain .O
.a f(/)= - ~
- g. i k E nM Rln- ,,
Sn O M d, , fgg) O # 1 ff/)df 2' =si .. 47) o O In order to evaluate the heat transfer coefficient, h, we consider the fact that, at least initially, the particle surface l superheat will be very high and, therefore, the heat transfer O coefficient is a combination of the film boiling and radiation components. According to Collier, p. 218, 'o if Y!~ > 2 0 h=h PB,C + 0.75 h r ' ' ' t 'O _ _ =_- __ ,
51
'O
.s l where
=27 4 lv b 'ff '
(f.f)
- o e
lab b54 MH
- f=l f/ f $ //fh_V/
O Tj-C )
$,=KG 7_ y;, J I'#N With r,= 0.00557 m, and U = 1.53 m/sec (assumed mean velocity 2
I and the relations do apply, o Vys_ > 2 With T m = 1500 C, Tsat = 120 C !O v = 1380 c = AT,,t (Vapor at 0.2MPa) k, = 0.6 W/m C 3 P. = 1.15 kg/m v O 6 J/kg i gg = 2.1842x39 C = 2000 J/kg'C O r
= 5.727x10
-8 jk 4 c = 0.7
.O
= /Z73.z g, _ ,
4 - iio.2 O
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.: / = /s rs./ -
fri) lO
52 'O This is a very large heat transfer coefficient. For compari-2 son, the prequench h in fuel reflood tests is only about 50 W/m *C.
'O If with this large h the particle surface superheat remains high enough to preclude nucleate boiling, the constant h assumption made in this analysis remains valid. The eigenvalues a n, the terms S , the sums, the surface superheat V(a,t) , the instantaneous heat flux q(t) , and At q have been numerically evaluated on a
,O computer (see Appendix I for listing). Results.are tabulated below as functions of time. 'O
- 0 .
O lO i O i
- O
,0 1
1
---.,-m -_, - -- , - - - - , , , . - - ----------mn - - - - - - - - - - - - - - - - , ,
53 O O Surface Time (sec) Superheat ( C) q \ [ q dt
) O O
0.1000000 1312.314 1815455. 181545.5 0.2000000 1281.450 1772758. 358821.3 0.3000000 1256.513 1738259, 532647.2 0.4000000 1234.596 1707940. 703441.2 0.5000000 1214.577 1680245. 871465.7 O 0.6000000 1195.879 1654378, 1036904. 0.7000000 1178.163 1629869. - 1199891. 0.8000001 1161.208 1606415. 1360532. 0.9000001 1144.864 1583805. 1518913. 1.000000 1129.022 1561889. 1675101. 1.100000 1113.602 1540557. 1829157. O 1.200000 1098.546 1519729. 198113o. 1.300000 1083.808 1499340. 2131064. . 1.400000 1069.355 1479344. 2278999 1.500000 1055.160 1459708. 242496?. 1.600000 1041.203 1440399. 2569009. 1.700000 1027.467 1421397. 2711149. O 1.800000 1o13.941 1402686. 2851418. 1.900000 1000.615 1384250. 2989843. 2.000000 987.4789 1366078. 3126450. 2.100000 974.5281 1348162. 3261266. 2.200000 961.7563 1330493. 3394316. 2.300000 949.1588 1313066. 3525622. 'O 2.400000 936.7314 1295874. 3655210. 2.500000 924.4709 1278913. 3783101. 2.600000 912.3738 1262178. 3909319. 2.700000 900.4371 1245664. 4033885. 2.799999 888.6584 1229370. 4156822. 2.899999 877.0350 1213290. 4278151. -() 2.999999 865.5648 1197422. 4397893. 3.099999 854.2451 1181762. 4516069. 3.199999 843.0742 1166309. 4632700. 3.299999 832.0494 1151057. 4747806. 3.399999 821.1696 1136006. 4861406. 3.499999 810.4319 1121151. 4973521. (3 3.599999 799.8349 1106491. 5084170. 3 699999 789.3767 1092024. 5193373. 3.799999 779.0555 1077745. 5301147. 3.899998 768.8689 1063653. 5407513. 3.999998 758.8158 1049745. 5512487. 4.099998 748.8943 1036020. 5616089. O ,0
54 lO Other results, shown in Appendix I, pertain to (a) larger size (1 inch diameter) particles, same heat transfer coefficient; (b) same (small) particles, and smaller heat transfer coefficient; 2 h = 150 W/m *C. The larger particles result in larger integrated heat flux. However, j y would now be smaller since the particle size appears in the denominator. Effects of smaller heat transfer coefficient (which is still 3 times larger than that observed in reflood experiments ahead of the quench front) are as exp,ected . 2.4.4 Steam Generation Rate O Let n,= number of particles contained within the submerged portion of the downcomer. If At is the time spent by the debris in the liquid region, see equations (52) and (53), then A 5 6t is d2 the volume of debris at any time in contact with water. That volume would contain a maximum of bdft bl particles /7z) and the total heat transfer surface AHT becomes ?O Ab ' 8' = 4Thb = A-(m') .. (71) (A-O Rate of total heat transferred to liquid 34 E a A ,/ = f, f - j N - tu)
- o
.O
O Heat balance for steam production
~O H,,,= W,[9,. fG-7;)+f,,]
Therefore, mass rate of steam produced
.5 Ad [4f .
- o L[tg (L-7;)4,)
hV and, since k4
- o $' p, '
- O')
j= Kn, f. . . (, ,) o n A<n u 7Mayi:, .ey. //r) 4/
- o
#* =
3ff# ' '
- A f. Di (L -7z) qd ] l)
o With r, = 0.00557 m 3 l p, = 1.15 kg/m O c p3 = 4185 (J/kg*C) T,,(-T3: 40*c 6 h = 2.1842x10 J/kg g fg
~4 l
K g7 = 1.99x10 [ q dt . . . [71) o
- O
, _ . - . _ _ . _ _ . , _ . . . , _ , - _ . , _ . _ , , _., ,m., , ._ ,, _,._._g, , ,.-.
56 O Substituting equation (52) into (53) O a . i. r aI / i- m i ,,9
/ / FZSj hence, At = 1.586 sec.
max bt
-6 q dt x10 j2 At o K HT j1 (m/sec) (sec) (h=1383 ) (eq. 78) (eq. 77) m C O
O O 0.02 1.27 2.08608 415.13 8.3026 0.04 1.0062 1.6755 333.4 13.337 0 0.06 0.7844 1.310 260.7 15.64 0.08 0.5954 1.0293 204.8 16.386 0.10 0.4324 0.7579 150.8 15.08 0 0.12 0.2905 0.5161 102.7 12.325 0.14 0.1657 0.2980 59.3 8.303 0.15 0.1089 0.1973 39.3 5.889 0 0.1709 0 0 0 0 This heat transfer determined relationship is plotted in Figure 2. Note that the assumption of no subcooling would increase D j y by only 7.7%. We shall next plot the hydrodynamically determined relation-ship between j y and j2 coming from the flooding correlation G tQ=(+G - (n) where from equation (55), v,= 36.26, /U = 6.02. Hence, , jy= (6.06-%)2 , ,. [fp) That relationship is plotted in Figure 3. 9
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.
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. 3-5 2 2..
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A_=-.k_-
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59 O Inspection of Figures 2 and 3 indicates that the heat trans-fer induced vapor flow lies below the vapor flow that would cause I)~ countercurrent flow limitation, even with the high heat transfer coefficient assumed in this analysis. This indicates that no flooding restriction will take place. O If, for whatever reasons, one would postulate some limitation to the magnitude of metal downflow--note that complete absence of j2 is not possible since, in that case, there would be no vapor 'O generation to inhibit downflow--one could easily visualize that heat transfer between the downflowing hot debris and the rela-tively thin downcomer pipe wall (3/8 inches) would quickly impair lO the pipe wall strength and the downcomer would collapse into the wetwell pool. Any vapor generation within the pool, due to the falling debris, would then vent into the wetwell's vapor space, O thus, again precluding flooding conduction from occurring. O O i l 'O O
- O
60 3. 2.5 Numerical Evaluation of Debris Transport in CRD Room The following ten cases were numerically evaluated. 3 Melt Number of Downcomer Cover Case Mass in Pressurized Gravity Ruptured Blocking Entry No. RPV (kg) Discharge Discharge CRD Tubes Until Melted D 1 200,000 Yes --- 5 No 2 200,000 .--- Yes 5 No 3 200,000 Yes --- 5 Yes (390 sec delay) 3 4 200,000 Yes --- 5 Yes (300 sec delay) 5 250,000 Yes --- 5 No 6 250,000 --- Yes 5 No 3 7 250,000 Yes --- 10 No 8 250,000 --- Yes 10 No 9 250,000 Yes --- 1 No 3 10 250,000 --- Yes 1 No Detailed listing of results and of the computer program, 3 CMELT, that incorporated the equations developed in this report are given in Appendix II. The most interesting result shared by all of these cases is '3 that nearly all of the discharged debris ends up in the wetwell. The only cases showing minute fractions of the discharged debris entering the drywell, through the manway, are Case Nos. 4, 5, and 3 7, with the computed fractions equa'l to 0.007, 0.005, and 0.001, respectively. These fractions are much smaller than those expected through the entrainment process discussed by Henry. He concluded O that in the case of the plant that does not feature downcomers 3
61
)
in the reactor cavity region, sufficient depth of molten debris could accumulate on the cavity room floor so that a significant fraction, up to 15%, could be entrained by hot gases discharged through the RPV ruptures following the melt discharge. Due to the four downcomers in Shoreham, the depth of debris D in the CRD room is never very high. Therefore, we expect less than 10% melt entrainment, reaching the drywell floor, and only for the pressurized discharge cases. J All cases were terminated when less than -inch-thickness of molten layer remained on the CRD room floor. That accounts for the remainder of the discharged material. J Other results of interest: a) Time to discharge the melt from RPV vessel 3 Assumed Number of CRD Tube Assumed Final Time to Ruptured Initial (Total) Complete (each 10 cm Mass of Break Discharge Pressurized Gravity diameter Melt Flow ^#*" 2 3 (sec) Discharge Discharge vertically) (kg) (m ) 5 X - 5 2 x10 5 0.556 28 - X 5 2 x10 5 0.449 5 3 5 X - 5 2.5x10 0.651 5 29 - X 5 2.5x10 0.526 5 4 X - 10 2.5x10 0.779 5 9 21 - X 10 2.5x10 0.681 5 10 X - 1 2.5x10 0.367 5 60 - X 1 2.5x10 0.302 O O
62 O
.b) Maximum depth of debris on CRD room floor Approximate
- 4) Case Maximum Depth Time of Maximum No. (cm) Depth (sec)-
1 59 4-2 16 22
.() -
3 59 4 4 59 4 5 59 4 o 6 18 23 7 71 3 8 21 17 O 9 51 9 10 13 30 0 It will be noticed that only in Case Nos. 3, 4, and 5 does the maximum depth approach the height of the manway sill (60.95 cm). Actually, the maximum debris depth must have briefly exceeded O the sill height because minute amount of discharge over the sill was computed for these cases. That fraction clearly depends on the assumptions regarding lO (a) when would the sill plate melt and/or (b) whether the sill plate could become dislodged prior to its melting. We have no information from which to judge the probability of the latter. l O 'O
- O
_ . _ ~ _ _ .
63 0 2.6 Brief Review of the In-Vessel and Ex-Vessel Explosion Probabilities O Those interested in an indepth analysis of this topic should refer to Henry's report cited earlier. We have noticed that SAI also dismissed consideration of such so-called " steam explosions" O in their PRA report. Our reasons for agreeing with their position are as follows: (a) In-vessel explosion *: 'O Not likely because the RPV bottom' head / lower plenum region is crowded with the control rods, their guide tubes, and other supporting structure. Coherent slump of core onto the bottom head does not appear possible and noncoherent immersions of molten chunks could not produce any serious explosions. Further-more, there appears to be no clear path for producing accelerating water slugs that could impact the RPV upper head wall and create missiles. (b) Explosion in the CRD room: The four downcomers located in that room guarantee that i l if any water were to collect on the CRD room floor prior to the l vessel wall failure, its depth could not exceed more than 3 , inches. This depth is too shallow to create steam explosions that I may endanger containment integrity. (c) Explosion in wetwell pool: The ratio: water volume to volume of melt reaching the wetwell floor, far exceeds those ratios which could not sustain steam explosions under laboratory conditions. In addition, the 'O
. 64
- O s
) noncoherent " dump," as calculated for the downcomer flow, is also ; not conducive to energetic explosions that may breach the con-
- O tainment, either due to missiles or shock waves.
'O .O 1 O 1 I, 10 .O !O i ? f !O 1 1
- O
'O s ,,-,--,-,w,, -
- __,,,w,. .---v---,,,p--, -wpr w-+ ,- - y e>w e,--,
-65 ;O 2.7 Overall Conclusions and Recommendations (1) It is our impression that the CRD cooling water INjec-O tion was not accounted for in the SAI analysis. Its impact would be to delay the time of the RPV breach.
(2) Our analysis indicates that most of the molten debris O discharged from the reactor vessel will end up, within about 5 minutes after the RPV rupture, in the wetwell pool. (3) Consequently, the amount of the debris reaching the
~O drywell floor, outside the pedestal area, would be so small that, when spread over the floor, it would be coolable by convective and radiative heat transfer to the containment atmosphere, and by .O conduction to concrete (neglecting presence of water) . In other words, it appears likely that the temperature of that melt fraction will quickly fall below that required for initiation of concrete O
attack. If that is true, the amount of noncondensible gases generated by the concrete interaction on the drywell may be very small. 30 (4) During the 5 minutes, or less, that it takes for the discharged melt to drain into the CRD room downcomers, the melt / concrete interaction can indeed take place, however, over the much LO smaller floor area. The CRD room wetwell floor area is nearly 20 times smaller than the drywell floor area outside the pedestal l room. l(3 . (5) During its transport to the wetwell floor, the molten debris surface temperature should be cooled to below that required i for significant metal / water reaction and for the concrete
!O jO
66 'O interaction. We could not address the subsequent heat balance for the wetwell water, steam production and the containment pressure O rise because the conclusions are strongly influenced by the mag- , nitude of heat leakage to and through the containment walls. We recommend that SAI be asked to repeat the MARCH cal-O culation, except, this time, to assume that about 90% of the released melt be transported directly to the wetwell pool, within 5 minutes after vessel rupture, to examine the effect on contain- .O ment pressure. We believe, however, that there will be no debris / concrete interaction on the wetwell floor due to low debris .O temperature. (6) In our opinion, the RPV failure would not come about by the failure of the vessel wall but, rather, through rupture / melt- -O through of one or more control rod guide tubes. That, however, plays a minor role in the overall scenario since all of the debris would be discharged in less than minute.
- O O
iO
- O O
67
.O 3.0 IMPACT OF RESULTS OF THIS STUDY ON SAI RESULTS CONCERNING ACCIDENT CONSEQUENCES Class I Scenario Expect very significant delay in the containment rupture time. 'If heat leakage to containment walls is significant, the containment may not rupture. It will certainly not rupture if containment sprays are operational. Need containment code analy-ses for quantification. Large decontamination factor is appro-priate for fission products (FP) released within the wetwell pool.
Class II Scenario O No significant impact, except for scrubbing action of FP within the pool. Class III Scenario O Same impact as for the class I scenario. Class IV Scenario I) No impact on containment failure (which occurs before vessel melt-through). However, the subsequent high pressure peak, around 120 minutes, may not occur because of very much smaller I) production of noncondensifale gases. This would significantly affect the transport of fission products through the ruptured containment. Containment code calculations are needed to quantify O the pressure history. Wetwell pool scrubbing of FP. l lO O
! 68 P
Class V Scenario Some details would change after the vessel melt-through. O l However, no significant impact since the' containment pressure i i remains low. FP decontamination within the wetwell pool.and much smaller production of noncondensible gases, would cause lower 'O i venting rates from containment through ruptured vessel (if flow passages not' blocked) to the secondary containment. O
- O i
i f l g 4 0 - f 1 i i -
69- !
.O i
4.0 LIST OF OTHER REFERENCES PERUSED DURING THIS STUDY J. G. Collier, Convective Boiling and Condensation, McGraw-10 Hill, 1972. G. B. Wallis, One-Dimensional Two-Phase Flow, McGraw-Hill, IO 1969. Carslaw and Jaeger, Conduction of Heat in Solids, Second 'O Edition, Oxford Press, 1959. '.O O .O 4 ) !O 'O i 10 t t
7e O APPENDIX I O Listing of Ccmputer Program HTPANS Developed for this Study, and Results O (a) HTRANS Listing O 0001 PROGRAM HTRANS 0002 DIMENSION ALFN(6),BB(9),CC(9) 0 0003 DATA CC/-1.0, -0.95, -0.9, -0.5, 0.0, 0.5, 1.0, 10., 100./ 0004 DATA BB/0., 0.3854, 0 5423, 1.1656, 1.5708, 1.8366, 2.20288.
- 2.8628, 3.1105/
C 0005 10 CONTINilE 0006 GTOT=0.0 0 0007 NITER =0 0008 TYPE 300 0009 300 FORMAT (1H1,//' ENTER H(W/M**2,DEG C) AND PARTICl.F RADTOS A(M)'/) 0010 ACCEPT *,H,A 0011 IF(H.EG.0)STOP 0013 DIF=4.167E-6 .O 0014 COND=22. - 0015 HP=H/COND 0016 AHP=A*HP 0017 C=AHP-1. 0018 DD 100 I=1,9 - 0019 IF(CC(I).GE.C)GO'TO 102 O 0021 100 CONTINilE 0022 102 CONTINUE 0023 BO=BB(I-1)+(BB(I)-BB(I-1))*(C-CC(I-1))/(CC(I)-CC(I-1)) 0024 .1 = 1 0025 105 CONTINilE 0026 NITER = NITER +1 0 0027 COTB=COS(BO)/ SIN (BO) 0028 BN=BO-(BO*COTB+C)/(C0TB*(1.-BO*COTB)-BO) 0029 IF(ABS ((BN-BO)/BO).LE.1.0E-5)GO TO 110 0031 *IF(NITER.GE.20)GO TO 400 0033 BO=BN 0034 GO TO 105 O O
7/
'O O
0035 400 TYPE 410 0036 410 FORMAT (' NUMBER OF ITFRATIONS MORE THAN 20') 0037 STOP 0038 110 ALFN(J)=BN/A 0039 NITER =0 0 0040 IF(J.EO.6)GO T,0 s20 0042 J=J+1 0043 BO=BN+3.1416 0044 GO TO 105 0045 120 CONTINllE 0046 TYPE 200,(AL.FN(I),1=1,6)
- O 0047 2o0 FORMAT (' EIGENVALUES Al.FN=',6F10.4/)
0048 TM0=1773. 0049 TSAT=393. 0050 V=THO-TSAT 0051 T=0.0 0052 DT=0.10
)
0053 1 T=T+DT 0054 SilMVS=0. 0055 SUM 01=0. 0056 SilM02=0.
'O c 0057 DO 2 J=1,6 0058 ALF=ALFN(J) 0059 ALFS=ALF*ALF 0060 AN=ALF*A 0061 ANS=AN*AN -O 0062 E=-DIF*ALFS*T 0063 EX=0.
i 0064 IF(-E.LE.50.)EX=EXP(E) l 0066 DEN =ALFS*(ANS+AHP*C) 0067 XK=2.*HP*V8EX*(ANS+C*C)* SIN (AN)/ DEN 0068 SUHVS=SilMVS+XK* SIN (AN) 0 0069 Sun 01=SiinvS 0070 SUMQ2= SUM 02+Al.F*XK*COS(AN) - 0071 2 CONTINUE C 0072 VS=SilHVS/A 0073 0=-COND/Af SilH02-SilH01/A) 0 0074 OTOT=0 TOT +a*DT 0075 TYPE *,TeVS,0,GTOT 0076 IF(T.GE.4.0)GO TO 10
- 0078 GO TO 1 0079 END O
l o
ns O (b) Additional Results Case 2: Heat transfer coefficient, h= 1383.4(W/sq.m -C) Particle radius = 0.0127 m (1 inch diam.) First column: Elapsed time (secs) Second column: Spherical particle surfacc temperature, deg C 3rd and 4th col.: q (W/m 2 ), and ff dl , respectively ENTER H(bl/M**2,DEG C) AND PARTICL.E RADIllS A(M) O 1383., 0.0127 EIGENU AL.tlES Al.FN= 112.6610 367.4561 614.3961 RA4.3471 1112.03RA Mus) Y('C) f ffM 1R1520.9 0.1000000 1312.516 1R15209. 0.2000000 1288.523 1782027. 359723.4 0 0.3000000 1268.529 1754375. 535161.1 0.4000000 1251.203 1730414. 70R202.4 0.5000000 1235.757 1709051. R79107.6 0.6000000 1221.699 1689610. , 104R069. 0.7000000 1208.712 1671649. 1215234. 0.8000001 1196.579 1654868. 13R0720. O 0.9000001 1185.144 1639054. 1544A26. 1.000000 1174.295 1624049. 1707031. 1 100000 1163.946 1609737. 1R6R004. 1 200000 1154.030 1596023. 2027607. 1.300000 1144.493 15R2R34. 2185R90. 1.400000 1135.294 1570111. 2342901. O 1.500o00 1126.395 1557804. 249R682. 1.600000 1117.74R 1545R73.- 2453269. 1.700000 1109.387 1534282. 2806697. 1.800000 1101.231 ' 1523002. 295R997. 1.900000 1093.281 1512007. 3110198. 2.000000 1085.521 1501276. 7260326. O 2.200000 1077.937 1490787. 3409404. 2.200000 1070.517 1480525. 3557457. 2.300000 1043.249 1470473. 3704504. 2.400000 1056.123 1460619. 3R50566.' 2.500000 1049.132 1450949. 3995461. 2.600000 1042.266 1441453. 4139806. O 2.700000 1035.518 1432121. 4283018. 1422944. 4425313,
'2.799999 1028.882 2.899999 . 1022.353 1413914. 4564704.
2.999999 1015.924 1405022. 4707206. 3.099999 1009.590 1396263. 4846833. 3.199999 1003.347 1387630. 4985596. O 3.299999 997.1914 1379116. 5123507. 3.399999 991.1183 1370716. 5260579. 3.499999 9R5.1243 1362427. 5396821. 3.599999 979.2057 1354242. 5532245. 3.699999 973.3600 1346157. 5666R61. 3.799999 967.5R40 133R169. 5800678. O 3.899998 961.8746 1330273. 5933705. 3.999998 956.2301 1322466. 4065952. 4.099998 950.6473 1314745. 6197426
73 O O case 3: neat transfer coefficient, h=1so w/m2-deg c Particle radius = 0.00557 m i O W bs* \ Y?0. . _ . . . . i 0.1000000 1372.410 205841.1 20586.11
. 0 2000000 1368.777 205315.9 41117.70 0.3000000 1365.763 204864.7 61604.17
() 0.4000000 1363.056 204457.3 82049.90 0 5000000 1360.534 204079.9 102457.9 0 6000000 1358.138 203720.3 . 122829.9 0.7000000 1355.829 203374.6 143167.4
- 0.8000001 1353.586 203037.7 163471.2 O.9000001 1351.392 202708.9 183742.0 0 1.000000 1349.235 202384.5 203980.5 1.100000 1347.107 202065.8 224187.1 l 1.200000 1345.001 201750.6 244362 1 1.300000 1342.914 201437.2 264505.9 l 1.400000 1340.840 201125.8 284618.5 1.500000 1338.779 200816.8 304700.2 O 1.600000 1336.727 200508.8 324751.0
- 1.700000 1334.683 200201.7 344771.;
1.800000 1332.646 199896.1 364760,8 1 900000 1330.614 199591.4 384720.9 2.000000 1328.589 199287.7 404648.8 2.100000 1326.568 198985.0 424547.3 2.200000 1324.551 198682.3 444415.5 - 9 2 300000 1322.538 198380.5 464253.6 i 2.400000 1320.529 198078.7 4840s1.4 2.500000 1318.524 197778.9 503839.3 l '2.600000 1316.522 197477.6 523587.1 l 2.700000 1314.524 197177.8 543304.8 2.799999 1312.528 196879.4 562992,8 ,() l 2.899999 1310.536 196580.5
- 582650.8 l 2.999999 1308.547 196281.6 602279.0
- 3.099999 1306.562 195983.7 621877.4 3.199999 1304 579 195686.3 641446.0 3.299999 1302.600 195390.3 660985.0
,3 3.399999 1300.623 195093.3 680494.3 3.499999 1298.649 194797.4 699974.1 3.599999 1296.679 194501.4 719424.2 3.699999 1294.712 194206.4 738844.8 [ 3.799999 1292.747 193912.3 758236.1 ! 3.899998 1290.786 193617.3 777597.8 Lg 3.999998 1288.828 193324.2 796930.3 4.099998 1286.872 193030.6 816233.3 l
- o .
7/ '
)
i APPDOIX II Idsting and Pesults, of Canputer Program QELT, to Ccrnpute Transport of
) !blten Df Is 0001 ~
PROGRAM CHELT_ ' C 0002 1 CONTINUE 0003 II=0
) 0004 IF(T.EG.O.) GO TO 304 0006 SUMSIL=SUNSIt.*8600./EMNO 0007 SUMWW=SUMWW*8600./EMMO 0008 TYPE 305,SUNSIL,SUMWW 0009 304 CONTINUE 0010 305 FORMAT (/' ACCUMULATED. FRACTION OF DEBRIS int'/ ) *' DRYWELL(OUTSIDE CRD ROUM)=',E15.5/
4' WETWELL P00L=,',E15.5) 0011 TYPE 1000 0012 2000 FORMAT (1H1//,' ENTER IPRESS,DT,TMSUMP,TMCOUR,TMSILL,NBRKS,
- RBRKO')
0013 ACCEPT *,IPRESS,DT,TMSUP,TMCOVR,TMSILL,NBRKS,RBRKO
) 0014 IF(IPRESS.L.T.0)STOP 0016 TYPE 1100 0017 1100 FORMAT (' ENTER INITIAL MOLTEN MASS IN RPV(KG)')
0018 ACCEPT *,EMMO 0019 IF(IPRESS.EO.1) TYPE 100 0021 ) 00?3 IF(IPRESS.EO.0) TYPE 110 TYPE 200 0024 T=0.0 0025 DTPRNT=1.0 0026 TPRINT=1.0 0077 HRPV=1.855*EMM0/2.Ef5 0020 EMM=EMMO
) 0029 HCRD=0.0 0030 SUNSIL=0.
0031 SUMWW=0. 0032 JITER=0 0033 2000 CONTINUE 0034 JITER=JITER+1 " 0035 HRPVS=HRPV*HRPV 0036 F=2.27*HRPVS-0.333*HRPVS*HRPV-3.7013E-5*EMH 0037 DF=5.54*HRPV-HRPVS 0038 HRPVN=HRPV-F/DF 0039 IF(ABS ((HRPVN-HRPV)/HRPV).LE.1.E-5)GO TO 2100 0041 IF(JITER.LT.50)GO TO 2050 0043 Ir(IPRESS.EG.0)STOP 0045 GO TO 20 0046 2050 CONTINUE 0047 HRPV=HRPVN l 0048 GO TO 2000 0049 2100 CONTINUE l
7T J 0050 IF(IPRESS.EO.1)GO TO 20 0052 SUM 0= SORT (HRPV) 0053 TMLTMN=THCOVR 0054 IF(TMSILL.L.T.THCOVR)TML.TMN=TMSILL 0056 10 , CONTINUE
) C 0057 T=TfDT 0058 IF(IPRESS.EO.0)GO TO 12 C----PRESSURI7ED DISCHARGE C
3 0060 GBRK=0. 0061 IF(HRPV.LT.1.0E-2)GO TO 20 0063 RBRK=RBRKO+0.0307*T 0064 R2=RBRK*RBRK 0065 ABRK=NBRKS*3.1416*R2 0066 OBRK=75.4*NBRKS*R2
] 0067 EMM=6.484E+5*NBRKS*T*(2.5E-3+1.535E-3*T+3.1416E-4*T*T) 0068 IF(EMM.GE.EMNO)0BRK=0. -
0070 IF(EMM.GE.EMMO)EMM=EMMO 0072 JITER=0 0073 EMM=EMMO-EMM 0074 GO TO 2000 0075 12 CONTINUE C C----GRAVITY DISCHARGE I C 0076 GBRK=0. 0077 IF(HRPV.LE.O.)GO TO 14
) 0079 0080 Y= SORT (HRPV)
SUMM= SUM 0+0.5*DT*(Y+YO) 00R1 RBRK=RBRKO+3.443E-3*SUMM 0082 R2=RBRK*RBRK ' 0083 ABRK=NBRKS*3.1416*R2 0084 TERM =4.43*NBRKS*DT/(Y*(5.54-HRPV)) 3 0085 HRPV=HRPV-TERM *R2 0086 IF(HRPV.LT.O.)HRPV=0.0 0088 YO=Y - 0089 SilM0=SUMM 0090 GBRK=8.35*NBRKS*R2*Y
- 0091 14 CONTINUE
) . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _
l l 0092 20 CONTINUE C
) C---MELT Fl.0W THROUGH DOWNCOMERS C
l 0093 ODWNC=0. 0094 IF(HCRD.L.E.O.114.AND.T.L.E.5.)GO TO 30 1 00Y6 IF(II.ED.1)GO TO 23 0098 TMSUP=T 0099 II=1 0100 23 CONTINUE i 0101 IF(T.GT.TMSUP)HTD=0.038*(1.-T/300.) 0103 IF(HTD.L.T.O.)HTD=0. 0105 IF(TMCOVR.GT.0.)GO TO 21 1 0107 GO TO 22
7/ O O 0108 21 CONTINUE () 0109 IF(T.GT.TMSUP+10..AND.T.LT.TMSUP+TMCOVR)GO TO 30 0111 IF(T.GE.TMSUP+TMCOVR)HTD=0. 0113 22 CONTINUE ' 0114 IF(HCRD-HTD.l.E.0.)GO TO 30 0116 ODWNC=22 62*(HCRD-HTD)**1.5 0117 30 CONTINUE C C---MELT FLOW PAST MANWAY SILL C O oits QSILL=0. 0119 IF(HCRD.LT.0.6096.AND.T.l.T.TMSIL.L)GO TO 40 0121 I F ( T . L.T . T MS II.L ) 0 S IL.l = 1. 8 * ( HC RD-0. 6 09 6 ) *
- 1. 5 0123 IF(T.GE.TMSILL)GSILL=1.8*HCRD**1.5 0125 40 CONTINllE 0126 SUMSit.=SUNSIL+GSILL*DT I) 0127 SilMWW=SUMWW+00WNC*DT C
C---NOW FIND THE HELT POOL LEVEL IN THE CRB ROOM C 0128 HCRD=HCRD+DT/23 13*(QBRK-GDWNC-GSILL.) 0129 IF(T.1.T.TPRINT)GO TO 42
) 0131 TYPE 300,T*,RBRK,ABRK,HRPV,HCRD,GBRK,GDWNC,OSII.t -
l 0132 IF(T.GE.30.)DTPRNT=10. 0134 IF(HCRD.t.E.0.0127.AND.T.GT.20.)GO TO 1 01.36 TPRINT=T+DTPRNT 0137 42 CONTINUE 1 8 GO To 10 O 0139 100 FORMAT (/' PRESSURIZED DISCHARGE CASE'/) 0140 110 FORMAT (/' GRAVITY DISCHARGE CASE'/) 0141 105 , FORMAT (' DT=',F5.3,' T MSilP , THCOVR, T MSIt.L e ' ,3F 10. 3,
*' NBRKS=',15,' RBRK0=',F10.5/)
0142 200 FORMAT (' T(SEC) RBRK(M) ABRK(M*M) HRPV(M) HCRD(M) O
- OBRK(M**3/SFc) oDWNC asit.L'/>
0143 300 FORMAT (F10.1,4F10.3,3F12.5) O e O o344 gwp i 1
79 O O - Ncrenclature for CMELT Input IPRESS= Indicator, =1 for pressurized discharge, =0 for gravity discharge O I7f = Integration time increments (secs)
'INSUP = Time to melt cover support columns. Not utilized frcrn input. 'ItCOVR= Time to melt /rsnove downcarer cover plates in CRD recrn 'INSIIL= Time to melt the manway sill NBRKS = Assumed number of ruptured CRD tubes RBRKO = Initial radius of each break (m)
Nanenclature for OELT Output O T = Elapsed time, secs, fran start of RPV rupture RBRK = Instantaneous radius (m) of each " hole" in RPV bottan head ABRK = Instantaneous, total break flow area (sq.m) HRPV = Height of molten debris pool within reactor vessel (m) HCRD QBBK
= Volumetric
= Height of molten flow ofdebris pool onvessel melt through CRD room floor rupture (m(mg/sec)
O ODKC = Volumetric flow of melt through 4 downcaners in CRD rocm OSILL = Volumetric flow of melt through CRD room manway 9 O O O O
w w w 1 Case 1. See page 60 in text for case description l ENTER IPRESS,DT,TMSUMP,TMCOUR,TMSILL,NBRKS, RBRKO 1, 0.5, 50., 0., 390., 5, 0.05 ENTER INITIAL MOLTEN MASS IN RPV(KG) 2.0E+5 PRESSURIZED DISCHARGE CASE T(SEC) RBRK(M) ABRK(M*M) HRPV(M) HCRD(M) nRRK(M**3/SEC) GDWNC OSILL 1.0 0.0$1 0.102 2.091 0.088 0.24552E+01 0.00000E+00 0.00000E+00 2.0 0.111 0.195 1.871 0.232 0.44786E+01 0.14894E+01 0.00000E+00 3.0 0.142 0.317 1.465 0.412 0.76125E+01 0.34131E+01 0.00000E+00 4.0 0.173 0.469 0 640 0.590 0.11257E+02 0.71587E+01 0.00000E+00 50 0.188 0.556 0.000 0.287 0.00000E+00 0.47241E+01 0.00000E+00 6.0 0.188 0.556 0.000 0.186 0.00000E+00 0.18573E+01 0.00000E+00 7.0 0.188 0.556 0.000 0.137 0.00000E+00 0.94989E+00 0.00000E+00 8.0 0.188 0.556 0.000 0.110 0.00000E+00 0 55922F+00 0.00000E+00 9.0 0.188 0.556 0.000 0.092 0.00000E+00 0.36016E+00 0.00000E+00 10.0 0.188 0.556 0.000 0.081 0.00000E+00 0.24705E+00 0.00000E+00 11.0 0.188 0.556 0.000 0.072 0.00000E+00 0.17763E+00 0.00000E+00 12.0 0.188 0.556 0.000 .0.066 0.00000E+00 0.13248E+00 0.00000E+00 13.0 0.188 0.556 0.000 0.061 0.00000E+00 0.10177Ef00 0.00000E+00 14.0 0.188 0.556 0.000 0.058 0.00000E+00 0.80105E-01 0.00000Ef00 15.0 0.188 0.556 0.000 0.055 0.00000E+00 0.64362E-01 0.00000E+00 N 16.0 0.188 0.556 0.000 0.052 0.00000E+00 0 57434E-01 0.00000E+00 17.0 0.188 0.556 0.000 0.050 0.00000E+00 0.43708E-01 0.00000E+00 18.0 0.188 0.556 0.000 0.049 0.00000E+00 0 347R9E-01 0.00000E+00 19.0 0.188 0.556 0.000 0.047 0.00000E+00 0.31341E-01 0.00000E+00
S Stu Y % Ncn Atax $Uwse & cut 20.0 0.188 0.556 0.000 0.046 0.00000E+00 0.26990E-01 0 00000E+00 21.0 0.188 0.556 0.000 0.045 0.00000E+00 0.23471E-01 0.00000E+00 22.0 0.188 0.556 0.000 0.044 0.00000E+00 0.20595E-01 0.00000E+00 23.0 0.188 0.556 0.000 0.043 0.00000E+00 0.18220E-01 0.00000E+00 24.0 0.1R8 0.556 0.000 0.043 0.00000E+00 0,.16241E-01 0.00000E+00 25.0 0.188 0.556 0.000 0.042 0.00000E+00 0.14579E-01 0.00000E+00 26.0 0.188 0.556 0.000 0.041 0.00000E+00 0.13173E-01 0.00000E+00 27.0 0.188 0.556 0.000 0.041 0.00000E+00 0 11975E-01 0.00000E+00 28.0 0.188 0.554 0.000 0.040 0.00000E+00 0.10949E-01 0.00000E+00 29.0 0.188 0.554 0.000 0.040 0.00000E+00 0.10064E-01 0.00000F+00 30.0 0.188 0.556 0.000 0.040 0.00000E+00 0.92973E-02 0.00000E+00 40.0 0.188 0.556 0.000 0.037 0.00000E+00 0.52524E-02 0.00000E+00 50.0 0.188 0.556 0.000 0.035 0.00000E+00 0 39006E-02 0.00000E+00 60.0 0.188 0.556 0.000 0.033 0.00000E+00 0.33615E-02 0.00000E+00 70.0 0.188 0.556 0.000 0.032 0.00000F+00 0.31274E-02 0.00000E+00 80.0 0.188 0.556 0.000 0.030 0.00000E+00 0.30214E-02 0.00000E+00 ' 0.029 0.00000E+00 0.29726E-02 0.00000E+00 90.0 100.0 0.188
'O.188 0.556 0.556 0.000 0.000 0.028 0.00000F+00 0.29498E-02 0.00000F+00 )
110.0 0.188 0.556 0.000 0.027 0.00000E+00 0.29392E-02 0.00000F+00 4 120.0 0.188 0.556 0.000 0.025 0.00000E+00 0.29342E-02 0.00000E+00 g 130.0 0.188 0.556 0.000 0.024 0.00000E+00 0.29319E-02 0.00000E+00 140.0 0.188 0.556 0.000 0.023 0.00000E+00 0.29308F-02 0.00000F+00 { 150.0 0.188 0.556 0.000 0.021 0.00000E+00 0.29303E-02 0.00000E+00 4 160.0 0.188 0.556 0.000 0.020 0.00000E+00 0.29300E-02 0.00000E+00 % 170.0 0.188 0.556 0.000 0.019 0.00000E+00 0.29299E-02 0.00000E+00 180.0 0.188 0.556 0.000 0.018 0 00000E+00 0.29299E-02 0.00000E+00 190.0 0.188 0.556 0.000 0.016 0.00000E+00 0 29298E-02 0.00000E+00 200.0 0.188 0.556 0.000 0.015 0.00000E+00 0.29298E-02 0.00000E+00 210.0 0.188 0.556 0.000 0.014 0.00000E+00 0.29298E-02 0.00000F+00 220.0 'O.188 .0.556 0.000 0.013 0.00000E+00 0.29298E-02 0.00000E+00 ACCUMilLATED FRACTION OF DEBRIS IN: DRYWELL(OllTSIDE CRD ROOM)= 0.00000E+00 WETWELL POOL = 0.98706E+00
l l l Case 2 1 i i I 4 ENTER IPRESS,DT,TMSilMP,TMCOVR,TMSILL,NBRKS, RBRKO
- 0,,,,,,,,,,,,,,
ENTER INITIAL MOLTEN MASS IN RPV(KG) l ,, GRAVITY DISCHARGE CASE T(SEC) RBRK(M) ABRK(M*M) HRPV(M) HCRD(M) GBRK(M**3/SEC) GDWNC OSII.L 1.0 0.059 0.055 2.178 0.009 0.21426E+00 0.00000E+00 0.00000E+00 2.0 0.044 0'.044 2.160 0.019 0.25187E+00 0.00000E+00 0.00000E+00 3.0 0.069 0.075 2.140 0.032 0.29198E+00 0.00000E+00 0.00000F+00 4.0 0.074 0.086 2.114 0.046 0.33439E+00 0.00000E+00 0.00000E+00 5.0 0.079 0.098 2.089 0.041 0.37885F+00 0.00000E+00 0.00000E+00 6.0 0.084 0.111 2.059 0.071 0.42509E+00 0.10141Ef00 0.00000E+00 7.0 0.089 0.124 2.024 0.084 0.47277E+00 0.18720E+00 0.00000E+00 8.0 0.094 0.139 1.986 . 0.095 0.52157F+00 0.27563E+00 0.00000E+00 9.0 0.099 0.153 1.944 0.105 0.57106E+00 0.34037E+00 0.00000E+00 10.0 0.104 0.148 1.898 0.113 0.62083Ef00 0.43902E+00 0.00000E+00 11.0 0.108 0.184 1.847 0.120 0.67037F+00 0.51132E+00 0.00000E+00 12.0 0.113 0.201 g 1.792 0.126 0.71917E+00 0.57792E+00 0.00000E+00 13.0 0.118 0.217 1.732 0.132 0.74662E+00 0.63964E+00 0.00000E+00 14.0 06122 0.234 1.667 0.137 0.81209E+00 0.69714E+00 0.00000E+00 , 15.0 0.127 0.252 1.598 0.142 0.85486E+00 0.75075E+00 0.00000E+00-I mu
1 I 4 b JAc hk. Mv Neto $3ep. kame kut. l 16.0 0.131 0.249 1.523 0.144 0.89414E+00 0 80051E+00 0.00000E+00 17 0 0.135 0.287 1.442 0.149 0.92906E+00 0.84616E+00 0.00000E+00 18.0 0.139 0.305 1.356 0.153 0.95862E+00 0.88721E+00 0.00000E+00 19 0 0.143 0.323 1.263 0.155 0.98168E+00 0.92296E+00 0.00000E+00 20.0 0.147 0.340 1.164 0.157 0.99491E+00 0.95250E+00 0.00000E+00 21.0 0.151 0.358 1.058 0.159 0.10027E+01 0.97473E+00 0.00000E+00 22.0 0.154 0.375 0.944 0.159 0.99701E+00 0.98828E+00 0.00000E+00 l 23.0 0.158 0.391 0.821 0.159 0.97714E+00 0.99148E+00 0.00000E+00 l 24.0 0.161 0.407 0 686 0.158 0.93917E+00 0 98211E+00 0.00000E+00
- 25.0 0.164 0.421 0.534 0.155 0.87676E+00 0.95713E+00 0.00000E+00 26 0 0.166 0.434 0.365 0.149 0.77771E+00 0.91177E+00 0.00000E+00 27.0 0 168 0.445 0.150 0.141 0 60950E+00 0.83715E+00 0.00000E+00 l 28.0 0 169 0.449 0.000 0.118 0.00000E+00 0.70817E+00 0.00000E+00 29 0 0.169 0.449 0.000 0.097 0.00000E+00 0.43909E+00 0.00000E+00 j
30.0 0.169 0.449 0.000 0.083 0 00000E+00 0.29311E+00 0.00000E+00 i 40.0 0.169 0.449 0 000 0.044 0.00000E+00 0 28922E-01 0.00000E+00 50.0 0 169 0.449 0.000 0.037 0.00000F+00 0.96451E-02 0.00000E+00 60.0 0.169 0.449 0.000 0.034 0.00000E+00 0.53561E-02 0.00000E+00 70 0 0.169 0.449 0.000 0.032 0 00000E+00 0.39395E-02 0.00000E+00 80 0 0.169 0.449 0.000 0.031 0.00000F.+00 0.33779E-02 0.00000E+00 90 0 0.169 0.449 0.000 0.029 0.00000E+00 0.31347E-02 0.00000E+00 100 0 0.169 0.449 0.000 0.028 0 00c)0E+00 0.30248E-02 0.00000E+00 110.0 0.169 0.449 0.000 0.027 9 00000E+00 0.29741E-02 0.00000E+00 120.0 0.169 0.449 0.000 0.025 0.00000E+00 0.29506E-02 0.00000F+00 130.0 0.169 0.449 0.000 0.024 0.00000E+00 0.29395E-02 0.00000E+00' 140.0 0.169 0.449 0 000 0.023 0.00000E+00 0.29344E-02 0.00000E+00 150 0 0.169 0.449 0.000 0.021 0 00000E+00 0.29319E-02 0.00000E+00 160.0 0.169 0.449 0.000 0.020 0.00000E+00 0.29308E-02 0.00000E+00 170.0 0.169 0.449 0.000 0.019 0 00000E+00 0.29303E-02 0.00000E+00 180.0 0.169 0.449 0.000 0.018 0.00000E+00 0.29300E-02 0.00000E+00 190.0 0.169 0 449 0.000 0.016 0.00000E+00 0.29299E-02 0.00000F+00 200.0 0.169 0.449 0.000 0.015 0.00000E+00 0.29298E-07 0.00000E+00 210.0 0.169 0.449 0.000 0.014 0.00000E+00 0.29298E-02 0.00000E+00 220.0 0.169 0.449 0.000 0.013 0.00000E+00 0.29298E-02 0.00000E+00 ACCtlMllLATED FRACTION OF DEBRIS IN: DRYWEl.L(OllTSIDE CRD ROOH)= 0.00000E+00 WETWELL POOL = 0.80383E+00
l l l l ENTER IPRESS,DT,TMSUMP,TMCOVR,TMSILL,NBRKS, RBRKO 1, 0.5, 50., 300., 390., 5, 0.05 l ENTER INITTAL MOLTEN MASS IN RPV(KG) 2.0E+5 Case 3 1 PRESSURIZED DTSCHARGE CASE T(SEC) RBRK(M) ABRK(M*M) HRPV(M) HCRD(M) GBRK(M**3/SEC) pHWNC GSILL l 1.0 0.081 0.102 2.091 0.08R 0.24552E+01 0 00000E+00 0.00000E+00 l , 2.0 0.111 0.195 1.871 0 235 0.46784E+01 0.13528E+01 0.00000F+00 3.0 0.142 0.317 1.445 0.413 0.74125E+01 0.34492E+01 0.00000E+00 4.0 0.173 0 469 0.640 0.591 0.11257E+02 0.71742E+01 0.00000E+00 l 5.0 0.188 0.554 0.000 0.287 0.00000E+00 0.47272E+01 0.Cn000E+00 t 6.0 0.188 0.556 0.000 0.184 0.00000EF00 0.18581E+01 0.00000E+00 I 7.0 0.188 0.556 0.000 0.137 0.00000F+00 0.95020E+00 0.00000E+00 8.0 0.188 0.556 0.000 0.110 0.00000E+00 0.55937E+00 0.00000E+00 l 9.0 0.180 0.556 0.000 0.092 0.00000E+00 0.36024E+00 0.00000E+00 10.0 0.188 0.556 0.000 0.081 0.00000E+00 0.24710E+00 0.00000E+00 l ,11.0 0.188 0.556 0.000 0.072 0.00000E400 0.17766E+00 0.00000F+00 12.0 0.188 0.554 0.000 0.066 0.00000E+00 0.13250E+00 0.00000E+00 . 13.0 0.188 0.556 0.000 0.066 0.00000E+00 0'.00000E+00 0.00000E+00 14.0 0.188 0.554 0.000 0.046 0.00000E+00 0.00000F+00 0.00000E+00 15.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 16.0 0.18R 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 17.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0 00000F+00 18.0 0.18R 0.554 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 19.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 20.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 21.0 0.18R 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 22.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00
- 23.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 C.00000E+00 24.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 j 25.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 26.0 0.188 0.556 0.000 0.044 0.00000E+00 0.00000E+00 0 00000E+00 27.0 0.188 0.556 0.000 0.044 0.00000E+00 0.00000E+00 0.00000E+00 28.0 0.188 0.556 0.000 0.046 0.00000F+00 0.00000E+00 0.00000E+00 29.0 0.188 0.554 0.000 0.064 0.00000E+00 0.00000E+00 0.00000E+00 30.0 0.188 0.554 0.000 0.066 0.00000E+00 0 00000E+00 0.00000E+00 40.0 0.188 0 556 0.000 0.064 0.00000E+00 0.00000E+00 0.00000E+00
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i Case 4 l l l ENTER IPRESS,DT,TMSilMP,THCOVR,TMSILL,NBRKS, RBRKO 1, 0.5, 50., 300., 300., 5, 0.05 ENTER INITIAL MOLTEN MASS IN RPU(KG) 2.0E+5 PRESSURIZED DISCHARGE CASE = T(SEC) RBRK(M) ABRK(M*M) HRPV(M) HCRD(M) GBRK(M**3/SEC) GDWNC OSIL L 1.0 0.081 0.102 2.091 0.088 0.24552E+01 0.00000E+00 0.00000E+00 2.0 0.111 0.195 1.871 0.232 0.46786E+01 0.14R94E+01 0.00000E+00 3.0 0.142 0.317 1.465 0.412 0.76125E+01 0.34131E+01 0.00000F+00 4.0 0.173 0.469 0.640 0.590 0.11257E+02 0 715R7E+01 0.00000E+00' 50 0.188 0.556 0.000 0.287 0.00000E+00 0.47241E+01 0.00000E+00 6.0 0.188 0.556 0.000 0.186 0.00000E+00 0.18573E+01 0.00000F+00 7.0 0.188 0.556 0.000 0.137 0.00000E+00 0.94989E+00 0.00000E+00 8.0 0.188 0.556 0.000 0.110 0.00000E+00 0 55922E+00 0.00000E+00 9.0 0.188 0.556 0.000 0.092 0.00000E+00 0.36016E+00 0.00000E+00' 10.0 0.18R 0.554 0.000 0.0R1 0.00000E+00 0.24705E+00 0.00000E+00 11.0 0.188 0.556 0.000 0.072 0.00000E+00 0.17763E+00 0.00000F+00 12.0 0.188 0.556 0.000 0.066 0.00000E+00 0.1324RE+00 0.00000E+00 13.0 0.188 0.556 0.000 0.064 0.00000E+00 0.00000E+00 0.00000E400 14.0 0.188 0.554 0.000 0.066 0.00000E+00 0.00000E+00 0.00000F+00 15.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 16.0 0.188 0.556 0.000 0.046 0.00000E+00 0.00000E+00 0.00000E+00 17.0 0.188 0.554 0.000 0.066 0.00000E+00 0.00000E+00 0.03000E+00 18.0 0.188 0.554 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 19.0 0.188 0.554 0.000 0.064 0.00000E+00 0.00000E+00 0.00000E+00 s 20.0 0.18R 0.554 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 % 21.0 0.18R 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000F+00 22.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00 23.0 0.1RR 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000F+00 24.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000Ff00 25.0 0.188 0.556 0.000 0.066 0.00000E+00 0.00000E+00 0.00000E+00
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Case 5 ENTER IPRESS,DT,TMSUMP,TMCOVR,TMSILL,NBRKS, RRRKO 1, 0.5, 50., 0., 300., 5, 0.05 ENTER INITIAL MOLTEN MASS IN RPV(KG) 2.5E+5 PRESSURIZED DISCHARGE CASE T(SEC) RRRK(M) ABRK(H*M) HRPV(M) HCRD(M) GBRK(M**3/SEC) GDWNC OSTLL 1.0 0.081 0.102 2.451 0.0RR 0 24552E+01 0.00000E+00 0.00000F+00 2.0 0.111 0.195 2.234 0.232 0.46784E+01 0.14R94E+01 0.00000E+00 3.0 0.142 0.317 1.853 0.412 0.76125E+01 0.34131E+01 0.00000E+00 4.0 0.173 0.469 1.200 0.590 0.11257E+02 0.71587E+01 0.00000E+00 5.0 0.204 0.651 0.000 0.427 0.00000E+00 0.11595E+02 0.32112E-01 6.0 0.204 0.651 0.000 0.239 0.00000E+00 0.31832E+01 0.00000F+00 7.0 0.204 0.651 0.000 0.164 0.00000E+00 0.14144E+01 0.00000E+00 8.0 0.204 0.651 0.000 0.125 0.00000E+00 0.76909E+00 0.00000E+00 9.0 0.204 0.651 0.000 0.102 0.00000E+00 0.47023E+00 0.00000E+00 10.0 0.204 0.651 0.000 0.0R7 0.00000E+00 0.31081E+00 0.00000E+00 11.0 0.204 0.651 0.000 0.077 0.00000E+00 0.21731E+00 0.00000E+00
~
12.0 0.204 0.651 0.000 0.070 0.00000E+00 0.15856E+00 0.00000E+00 13.0 0.204 0.651 0.000 0.064 0.00000E+00'O.11966E+00 0.00000E+00 14.0 0.204 0.651 0.000' O.060 0.00000E+00 0.92808E-01 0.00000E+00 15.0 O.204 0.651 0.000 0.056 0.00000E+00 0 73645E-01 0.00000E+00 16.0 0.?04 0.651 0.000 0.054 0.00000E+00 0.595R2E-01 0.00000E+00 17.0 0.204 0.651 0.000 0.052 0.00000E+00 0.49016E-01 0.00000E+00 18.0 0.204 0.651 0.000 0.050 0.00000E+00 0.40918E-01 0.00000E+00 19.0 0.204 0.651 0.000 0.048 0.00000E+00 0 34602E-01 0.00000E+00 S 20.0 0.204 0.651 0.000 0.047 0.00000E+00 0.29601E-01 0.00000E+00 21.0 0.204 0.651 0.000 0.046 0.00000E+00 0.255RRE-01 0.00000E+00 22.0 0.204 0.651 0.000 0.045 0.00000E+00 0.22329E-01 0.00000F+00 23.0 0.204 0.651 0.000 0.044 0.00000E+00 0>.19654E-01 0.00000E+00
O O O O O O O O O O O
~
Mc Nog S8& So w & cut 24.0 0.204 0.651 0.000 0.043 0.00000E+00 0.1743RE-01 0.00000E+00 25.0 0.204 0.651 0.000 0.042 0.00000E+00 0 15584F-01 0.00000E+00 26.0 0.204 0.651 0.000 0.042 0.00000E+00 0 14026E-01 0.00000E+00 27.0 0.204 0.651 0.000 0.041 0.00000E+00 0.12703E-01 0.00000F+00 28.0 0.204 0.651 0.000 0.041 0.00000E+00 0.11573E-01 0.00000E+00 29.0 0.204 0.651 0.000 0.040 0.00000F,+00 0.10603E-01 0.00000E+00 30.0 0.204 0.651 0.000 0.040 0.00000E+00 0.97644E-02 0.00000F+00 40.0 0.204 0.651 0.000 0.037 0.00000E+00 0.53913F-02 0.00000E+00 50.0 0.204 0.651 0.000 0.035 0.00000E+00 0.39527F-02 0.00000E+00 60.0 0.204 0.651 0.000 0.033 0.00000E+00 0.33834E-02 0.00000F+00 70.0 0.204 0.651 0.000 0.032 0.00000E+00 0.31371F-02 0.00000F+00 80.0 0.204 0.651 0.000 0.030 0.00000E+00 0.30259E-02 0.00000E+00 90.0 0.204 0.651 0.000 0.029 0.00000E+00 0.29747E-02 0.00000F+00 100.0 0.20A 0.651 0.000 0.028 0.00000E+00 0.29508E-02 0.00000E+00 110.0 0.204 0.651 0.000 0.027 0.00000E+00 0 29394E-02 0.00000E+00 120.0 0.204 0.651 0.000 0.025 0.00000E+00 0.29344E-02 0.00000E+00 130.0 0.204 0.651 0.000 0.024 0.00000E+00 0.29320E-02 0.00000E+00 140.0 0.204 0.651 0.000 0 023 0.00000E+00 0.29308E-02 0.00000E+00 150.0 0.204 0.651 0.000 0.021 0.00000F+00 0 29303E-02 0.00000E+00 160.0 0.204 0.651 0.000 0.020 0.00000E+00 0.29300E-02 0.00000E+00 170.0 0.204 0.651 0.000 0.019 0.00000E+00 0.29299E-02 0.00000E+00 180.0 0.204 0.651 0.000 0.018 0.00000F+00 0.29299E-02 0.00000E+00 190.0 0.204 0.651 0.000 0.016 0.00000F+00 0.29298E-02 0.00000E+00 200.0 0.204 0.651 0.000 0.015 0.00630E+00 0.29298E-02 0.00000E+00 210.0 0.204 0.651 0.000 0.014 0.00000E+00 0.29298E-02 0.00000E+00 220.0 0.204 0.651 0.000 0.013 0.00000E+00 0.29298E-02 0.00,000E+00 ACCllMllLATED FRACTION OF DEBRIS IN: DRYWElL(OllTSIDE CRD ROOM)= 0.55232E-03 g WETWELL POOL = 0.10186E+01 d
Casa 6. ENTER TPRESS,DT,TMSilMP,TMCOVR,TMSILL,NBRKS, RERKO 0,,,,,,,,,,,,,, ' ENTER INITIAL MOLTEN MASS IN RPV(KG) GRAVITY DISCHARGE CASE T(SEC) RBRK(H) ABRK(M*M) HRPV(M) HCRD(M) OHRK(M**3/SEC) RDWNC QSILL , 1.0 0.040 0.054 2.537 0.010 0.23945F+00 0.00000E+00 0.00000E+09 2.0 0.065 0.067 2.518 0.022 0.28427E+00 0.00000E+00 0.00000F40. 3.0 0.071 0.079 2.495 0.036 0.33235E+00 0.00000E+00 0.00000E400 4.0 0.074 0.092 2.470 0.052 0.38343E+00 0.00000E+00 0.00000E+00 5.0 0.082 0.105 2.440 0.070 0.43724E+00 0.00000E+00 0.00000Ef00 6.0 0.087 0.119 2.406 0.080 0.49345E+00 0.15156E+00 0.00000E+00 7.0 0.092 0.134 2.348 0.094 0.55168E+00 0 25282E+00 0.00000E+00 8.0 0.098 0.150 2.326 0.105 0.41153E+00 0.35409E+00 0.00000E+00 9.0 0.103 0.167 2.280 0.115 0.67253E+00 0.45014E+00 0.00000E+00 10.0 0.108 0.184 2.228 0.124 0 73417E+00 0.53950E+00 0.00000E+00 11.0 0.113 0 202 2.172 0.132 0.79588E+00 0.42253E+00 0.00000E+00 12 0 0.118 0.220 2.111 0.138 0.85705E+00 0.70020E+00 0.00000E+00 13.0 0.123 0.239 2.044 0.145 0.91702E+00 0 77342E+00 0.00000E+00 l l 14.0 0.128 0.259 1.973 0.151 0.97503E+00 0.84279E+00 0.00000E+00 15 0 0.133 0.279 1'.896 0.156 0.10303E+01 0.90854E+00 0.00000E+00 16.0 0.138 0.299 1.813 0.161 0.10819E+01 0.97056E+00 0.00000F+00 i 17.0 0.143 0.319 1.725 0.165 0.11290E+01 0.10284E+0! 0.00000E+00 l 18.0 0.147 0.340 1.631 0.169 0.11703E+01 0.10815E+01 0.00000E+00 0.173 0.12048E+01 0.11290E+01 0.00000E+00 19.0 0 151 0.360 1.530 ! 20.0 0.156 0.381 1.423 0.174 0.12310E+01 0.11698E+01 0.00000E+00 %
- 21.0 0.160 0.401 1.309 0.178 0.12473E+01 0.12028E+01 0.00000E+00
O O O O O O O O O O O i 0 &a 4ex H, _. ,, , Hm Reex Gew Osnt 22.0 0.144 0.421 1.187 0 179 0.12517E+01 0.12265E+01 0.00000E+00 23.0 0.168 0.441 1.057 0.179 0.12d19E+01 0.12392E+01 0.00000E+00 24.0 0.171 0.460 0.916 0.179 0.12143E+01 0.12387E+01 0.00000E+00 i 25.0 0.174 0.478 0.763 0.177 0.116A3E+01 0.12224E+01 0.00000F+00 ) 26.0 0.177 0.494 0.594 0.173 0.10840E+01 0.11843E+01 0.00000E+00
- 27.0 0.100 0.509 0.401 0.166 0.95804C+00 0.11245E+0'1 0 00000E+00
! 28.0 0.182 0.521 0.159 0.156 0.74502E400 0.10252E+01 0'.00000E+00
- 29.0 0.183 0.526 .0.000 0.128 0.00000E+00 0.85493E+00 0.00000E+00 30.0 0.183 0.526 0.000 0.103 0.00000E+00 0.51308E+00 0.00000E+00
! 40.0 0.183 0.526 0.000 0.046 0.00000E400 0.35694E-01 0.00000E+00 50.0 0.183 0.526 0.000 0.038 0.00000E+00 0.10777E-01 0.00000E+00 60.0 0.183 0.526 0.000 0.034 0.00000E+00 0.56827E-02 0.00000E+00 70.0 0.183 0.526 0.000 0.032 0.00000E+00 0.40605E-02 0.00000F400 i 80.0 0.183 0.526 0.000 0.031 0.00000E+00 0.34286E-02 0.00000E+00 90.0 0.183 0.524 0.000 0.029 0.00000E+00 0.31572E-02 0.00000E+00 100.0 0 183 0.526 0.000 0.028 0.00000E+00 0.30351E-02 0.00000E+00 110.0 0 183 0.524 0.000 0.027 0.00000E+00 0.29789E-02 0.00000E+00 120.0 0.183 0.526 0.000 0.025 0.00000E+00 0.29528E-07 0.00000E+00 130.0 0.183 0.524 0.000 0.024 0.00000E+00 0.29406E-02 0.00000E+00 140.0 0.183 0.526 0.000 0.023 0.00000E+00 0.79349E-02 0.00000E+00 150.0 0.183 0.526 0.000 0.021 0.00000E+00 0.29322E-02 0 00000E+00 160.0 0.183 0.526 0.000 0.020 0.00000E+00 0.29309E-02 0.00000E+00 170.0 0.183 0.526 0.000 0.019 0.00000E+00 0.29303E-02 0.00000F+00 180.0 0.183 0.526 0.000 0.018 0.00000E+00 0.29301E-02 0.00000E+00 190.0 0.183 0.526 0.000 0.016 0.00000E+00 0.29299E-02 OgC0000E+00 200.0 0.183 0.526 0.000 0.015 0.00000E+00 0.29299E-02 0.00000E+00 210.0 0.183 0.526 0.000 0.014 0.00000E+00 0.29298E-02 0.00000E+00 220.0 0.183 0.526 0.000 0.013 0.00000E+00 0.29298E-02 0.00000E+00 ACCtJMilLATED FRACTION OF DEBRIS IN! DRYWELL(OllTSIDE CRD ROOM)= 0.00000E+00 WETWELL PDOL= 0.82247E+00
i O O O O O O O O O O O-i l I Cas3 7 i i ENTER. IPRESS,DT,TMSUMP,TMCOVR,TMSILL,NBRKS, RBRKO 1, 0.5, 50., 0., 300., 10, 0.05 ENTER INITIAL MOLTEN MASS IN RPV(KG) 2.5E+5 PRESSURIZED DISCHARGE CASE T(SEC) RBRK(M) ABRK(M*M) HRPV(M) HCRD(M) QBRK(M**3/SEC) DDWNC OS Il. L
- 1.0 0.081 0.205 2.349 0.176 0.49104E+01 0.00000E+00 0.00000E+00 2.0 0.111 0.390 1.913 0.432 0.93571E+01 0.29200E+01 0.00000F+00 3.0 0.142 0.634 1 046 0.711 0.15225E+02 0.88599E+01 0.00000E+00 4.0 0.157 0.779 0.000 0.315 0.00000E+00 0.57621 E+0L 0.00000E+00 5.0 0.157 0.779 0.000 0.198 0.00000E+00 0.2114RE+01 0.00000F+00 6.0 0.157 0.779 0.000 0.144 0.00000E+00 0.10481F+01 0-00000F+00 7.0 0.157 0.779 0.000 0.114 0.00000E+00 0.40570E+00 0.00000E+00 8.0 0.157 0.779 0.000 0.095 0:00000E+00 0.38525E+00 0.00000E+00 9.0 0.157 0.779 0.000 0.082 0.00000E+00 0.2618HE+00 0.00000E+00 10.0 0.157 0.779 0.000 0.074 0.00000E+00 0.18699E+00 0.00000E+00 11.0 0.157 0.779 0.000 0.067 0.00000E+00 0.13871E400 0.00000E+00 12.0 0.157 0.779 0.000 0.062 0.00000E+00 0.1060RE+00 0.00000E+00 13.0 0.157 0.779 0.000 0.058 0.00000F+00 0.83189E-01 0.00000E+00 14.0 0.157 0.779 0.000 0.055 0.00000E+00 0.66630E-01 0.00000F+00 15.0 0.157 0.779 0.000 0.053 0.00000E+00 0.54340E-01 0.00000E GD 16#0 0.157 0.779 0.000 0.051 0.00000F+00 0.45017E-01 0.00000E+00 17.0 0.157 0.779 0.000 0.049 0.00000E+00 0.37812E-01 0.00000E+00 18.0 0.157 0.779 0.000 0.04R 0.00000E+00 0.3215tF-02 0.00000E+00 19.0 0.157 0.779 0.000 0.046 0.00000E+00 0.27640E-01 0.00000E+00 20.0 0.157 0.779 0.000 0.045 0.00000E+00 0.24000E-01 0.00000E+00 21.0 0.157 0.779 0.000 0.044 0.00000E+00 0.21029E-01 0.00000E+00 22.0 0.157 0.779 0.000 0.044 0.00000E+00 0.1R5 ROE-01 0,00000E+00
84x A SAd S CAb kgRg 0h w hu. 23.0 0.157 0.779 0.000 0.043 0.00000E+00 0.16542E-01 0.00000E+00 24.0 0 157 0.779 0.000 0.042 0.00000E+00 0.14833E-01 0.00000E+00 25.0 0.157 0.779 0.000 0.042 0.00000E+00 0.133RRE-01 0.00000E+00 26.0 0 157 0.779 0.000 0.041 0.00000E+00 0 12159E-01 0.00000F+00 27.0 0.157 0.779 0.000 0.041 0.00000E+00 0.11107E-01 0.00000E+00 28.0 0.157 0.779 0.000 0.040 0.00000E+00 0.10200E-01 0.00000E+00 29.0 0.157 0.779 0.000 0.040 0.00000E+00 0.9415RE-02 0.00000E+00 30.0 0 157 0.779 0.000 0.039 0.00000E+00 0.87333E-02 0.00000E+00 40.0 0.157 0.779 0.000 0.037 0.00000E+00 0.50811E-07 0.00000E+00 50.0 0.157 0.779 0.000 0.035 0.00000E+00 0.3R354E-02 0.00000F+00 60.0 0.157 0.779 0.000 0.033 0.00000E+00 0.3333RE-02 0.00000E+00 70.0 0.157 0.779 0.000 0.032 0.00000E+00 0.31150E-02 0.00000F+00 80.0 0.157 0 779 0.000 0.030 0.00000E+00 0.30157E-02 0.00000E+00 90.0 0.157 0.779 0.000 0.029 0.00000E+00 0.29499E-02 0.00000E+00 1 100.0 0 157 0.779 0.000 0.02R 0.00000E+00 0.294R6E-02 0.00000E+00 110.0 0.157 0.779 0.000 0.027 0.00000E+00 0.293R6E-02 0.00000E+00 120.0 0 157 0.779 0.000 0.025 0.00000E+00 0.29339E-02 0.00000E+00 130 0 0 157 0.779 0.000 0.024 0.00000E+00 0.29317E-02 0.00000E+00 140 0 0.157 0 779 0.000 0.023 0.00000E+00 0.29307E-02 0.00000E+00 150.0 0 157 0.779 0.000 0.021 0.00000E+00 0 29302E-02 0.00000E+00 160.0 0.157 0.779 0.000 0.020 0.00000E+00 0.29300E-02 0.00000E+00 170.0 0.157 0 779 0.000 0.019 0.00000E+00 0.29299E-02 0.00000E+00 180.0 0.157 0.779 . 0.000 0.01R 0.00000E+00 0.29299E-02 0.00000E+00 190 0 0.157 0.779 0.000 0.016 0.00000E+00 0.29298E-02 0.00000E+00 200.0 0.157 0.779 0.000 0.015 0.00000E+00 0.29298E-02 0.00000E+00 210.0 0.157 0.779 0.000 0.014 0.00000F+00 0.29298E-02 0.00000E400 220 0 0.157 0.779 0.000 0 013 0.00000E+00 0.29298E-02 0.00000E+00 ' l l ACCllMllLATED FRACTION OF DEBRIS IN' DRYWELL(OllTSIDE CRD ROOM)= 0.99300E-03 , y WETWELL POOL = 0.R7961E+00
Case 8 ENTER IPRESS,DT,TMSUMP,TMCOVR,TMSILL,NBRKS, RPRKO 0,re,,,,,,,,,, ENTER INITIAL MOL. TEN MASS IN RPU(KG) GRAVITY DISCHARGE CASE - T(SEC) RBRK(H) ABRK(M*M) HRPV(M) HCRD(M) OBRK(M**3/SEC) GDWNC OS Il. L
~
1.0 0.060 0.113 2.521 0.020 0.47830E+00 0.00000E+00 0.00000F+00 2.0 0.065 0.135 2.483 0.043 0.56555E+00 0.00000E+00 0.00000E+00 3.0 0.071 0.158 2.439 0.071 0.65775E+00 0.00000E+00 0.00000E+00 4.0 0.076 0.183 2.387 0.102 0.75393E+00 0.00000E+00 0.00000E+00 5.0 0.082 0.209 2.329 0.120 0.85296E+00 0.81938E+00 0.00000E+00 6.0 0.087 0.237 2.263 0.135 0.95357E+00 0.42014E+00 0.00000E+00 7.0 0.092 0.266 2.189 0.148 0 10543E+01 0.76868E+00 0.00000E+00 8.0 0.097 0.294 2.107 0.159 0.11536E+01 0.90487E+00 0.00000E+00 9.0 0.102 0.327 2.017 0.169 0.12496E+01 0.10300E+01 0 00000E+00
.1 O' . 0 0.107 0.359 1.918 0.178 0.13404E+01 0.11453E+01 0.00000E+00 11.0 0.112 0.392 1.810 0.186 0.14236E+01 0.12512E+01 0.00000E+00 12.0 0.116 0.425 1.693 0.192 0.14968E+01 0.13472E+01 0.00000E+00 13.0 'O.121 0.459 1.566 0.198 0.15569E+01 0.14320E+01 0.00000E+00 14.0 0.125 0.492 1.428 0.202 0.16008E+01 0.15034E+01 0.00000E+00 15.0 0.129 0.525 1.279 0.206 0.16241E+01 0.15588E+01 0.00000E+00 16.0 0.133 0.557 1.117 0.207 0.16218E+01 0 15945E+01 0.00000F+00 17.0 0.137 0.588 0.941 0.207 0.15865E+01 0.16061E+01 0.00000E+00 18.0 0.140 0.617 0.745 0.204 0.15074E+01 0.15875E+01 0.00000F+00 19.0 0.143 0.643 0.521 0.198 0.13647E+01 0.15296E+01 0.00000E+00 20.0 0.146 0.666 0.249 0.187 0.11103E+01 0.14165E+01 0.00000E+00 21.0 0.147 0.681 0.000 0.161 0.47266E+00 0.12077E+01 0.00000E+00 22.0 0.147 0.681 0.000 0.123 0.00000E+00 0.76048E+00 0.00000E+00 @
23.0 0.147 0.681 0.000 0.100 0.00000E+00 0.46586E+00 0.00000E+00 24.0 0.147 0.681 0.000 0.085 0.00000E+00 0.30835E+00 0.00000F+00
y - __ _ _ g . O ~U U O O l I l ifK bfg Nm hgg WM UK 25.0 0.147 0.681 0.000 0.075 0.00000E+00 0.21580E+00 0.00000E+00 26.0 0.147 0.681 0.000 0.068 0.00000E+00 0.15758E+00 0.00000E+00 27.0 0.147 0.681 0.000 0.062 0.00000E+00 0.11899E+00 0.00000F+00 28.0 0.147 0.681 0.000 0.058 0.00000E+00 0.92342E-01 0.00000E+00 29.0 0.147 0.681 0.000 0.055 0.00000E+00 0.73307E-01 0.00000E+00 30.0 0.147 0.681 0.000 0.052 0.00000E+00 0.59330E -01 0.00000F+00 40.0 0.147 0.681 0.000 0.040 0.00000E+00 0.13994E-01 0.00000E+00 50.0 0.147 0.681 0.000 0.036 0.00000E+00 0.65376E-02 0.00000E+00 60.0 0.147 0.681 0.000 0.034 0.00000Et00 0.43650E-02 0.00000E+00 70.0 0.147 0.681 0.000 0.032 0.00000E+00 0.35534E-02 0.00000E+00 80.0 0.147 0.681 0.000 0.031 0.00000E+00 0.32123E-02 0.00000E+00 90.0 0.147 0.681 0.000 0.029 0.00000E+00 0.30602E-02 0.00000F+00 100.0 0.147 0.681 0.000 0.028 0.00000EF00 0.29905E-02 0.00000E+00 110.0 0.147 0.681 0.000 0.027 0.00000E+00 0.29582E-02 0.00000E+00 . 120.0 0.147 0.681 0.000 0.025 0.00000E+00 0.29431E-02 0.00000E+00
.130.0 0.147 0.681 0.000 0.024 0.00000E+00 0.29360E-02 0.00000E+00 140.0 0.147 0.681 0.000 0.023 0.00000E+00 0.29327E-02 0.00000E+00 150.0 0.147 0.681 0.000 0.021 0.00000E+00 0.29312E-02 0.00000E+00 160.0 0.147 0.681 0.000 0.020 0.00000E+00 0.29304E-02 0.00000F+00 170.0 0.147 0.681 0.000 0.019 0.00000E+00 0.29301E-02 0.00000E+00 180.0 0.147 0.681 0.000 0.018 0.00000E+00 0.29299E-02 0.00000E+00 190.0 0.147 0.681
- 0.000 0.016 0.00000E+00 0.29299E-02 0.00000E+00 200.0 0.147 0.681 0.000 0.015 0.00000E+00 0.29298E-02 0.00000E+00 210.0 0.147 0.681 0.000 0.014 0.00000E+00 0.29298E-02 0.00000E+00 220.0 0.147 0.681 0.000 0.0,13 0.00000E+00 0.29298E-02 0.00000E+00 ACCUMULATED FRACTION OF DEBRIS IN:
DRYWELL(OUTSIDE CRD ROOM)= 0.00000E+00 WETWELL POOL = 0.83220E+00 3
Case 9.
- ENTER IPRESS,DT,TMSUMP,TMCOVR,TMSIt.L,NBRKS, RBRKO 1, 0.5, 50., 0., 300., 1, 0.05 ENTER INITIAL MOLTEN MASS IN RPV(KG) 2.5E+5 PRESSURIZED DISCHARGE CASE T(SEC) RBRK(M) ABRK(M*M) HRPV(M) HCRD(M) QBRK(M**3/SEC) GDWNC OSILL 1.0 0.081 0.020 2.532 0.018,0.49104E+00 0.00000Ef00 0.00000E+00 2.0 0.111 0.039 2.4R9 0.053 0.93571E+00 0.00000E+00 0.00000E+00 3.0 0.142 0.063 2.413 0.112 0.15225E+01 0.00000E+00 0.00000E+00 4.0 0.173 0.094 2.297 0.175 0.22514E+01 0.12131E+01 0.00000F+00 5.0 0.204 0.130 2.132 0.241 0.31225E+01 0.15887E+01 0.00000E+00 6.0 0.234 0.172 1.904 0.307 0.41357E+01 0.26050E+01 0.00000E+00 -
7.0 0.265 0.220 1.599 0.373 0.52910E+01 0.37704E+01 0.00000F+00 8.0 0 296 0.275 1.165 0.439 0.65884E+01 0.5064RE+01 0.00000E+00' 9.0 0.326 0.334 0.328 0.505 0.80280F+01 0.64872E+01 0.00000E+00 - 10.0 0.342 0.367 0.000 0.263 0.00000E+00 0.39353E+01 0.00000F+00 11.0 0.342 0.367 0.000 0.175 0.00000E+00 0.16410E+01 0.00000E+00 12.0 0.342 0.367 0.000 0.131 0.00000F+00 0.86341E+00 0.00000E+00 13.0 0.342 0.347 0.000 0.106 0.00000E+00 0.51735E400 0.00000E+00 14.0 0.342 0.367 0.000 0.090 0.00000E+00 0.33717E+00 0.00000E+00 15.0 0.342 0.347 0.000 0.078 0.00000Ef00 0.23329E+00 0.00000E+00 16.0 0.342 0.367 0.000 0.071 0.00000E+00 0.148R5E+00 0.00000E+00 17.0 0.342 0.367 0.000 0.065 0.00000E+00 0.12660E+00 0.00000E+00 18.0 0.342 0.367 0.000 0.060 0.00000Ef00 0.97677E-01 0.00000E+00 19.0 0.342 0.367 0 000 0.057 0.00000E+00 0.77164E-01 0.00000F+00 l 20.0 0.342 0.367 0.000 0.054 0.d0000E+00 0.62191E-01 0.00000E+00 l 21.0 0 342 0.367 0.000 0.051 0.00000E+00 0.50994E-Ot 0.00000E+00 22.0 0.342 0.347 0.000 0.050 0.00000E+00 0.42445E-01 0.0000JF+00 Q ! 23.0 0.342 0.347 0.000 0.048 0.00000E+00 0.35R01E-01 0.00000E+00 l
Stx 8k '! , NMb GSU 4=*W &M 24.0 0.342 0.367 0.000 0.047 0.00000E+00 0.30556E-01 0.00000E+00 25.0 0.342 0.367 0'.000 0.045 0.00000E+00 0.26358E-01 0.00000E+00 26.0 0.342 0.367 0.000 0.044 0.00000Ef00 0 22957E-01 0.00000E+00 27.0 0.342 0.347 0.000 0.043 0.00000E+00 0.20172E-01 0.00000E+00 28.0 0.342 0.367 0.000 0.043 0.00000E+00 0.17869E-01 0.00000F+00 L29.0 0.342 0.367 0.000 0.042 0.00000E+00 0 15947E-01 0.00000E+00 30.0 0.342 0.367 0.000 0.041 0.00000E+00 0.14331E-01 0.00000E+00 40.0 0.342 0.367 0.000 0.037 0.00000E+00 0.66212E-02 0.00000E+00 50.0 0.342 0.367 0.000 0.035 0.00000E+00 0.43939E-02 0.00000E+00 60.0 0.342 0.367 0.000 - 0.033 0.00000E+00 0.35453E-02 0.00000E+00
- 70.0 0.342 0.367 0.000 0.032 0.00000E+00 0.32174E-02 0.00000E+00 00.0 0.342 0.347 0.000 0.030 0.00000E+00 0.30625E-02 0.00000E+00 90.0 0.342 0.367 0.000 0.029 0.00000E+00 0.29916E-02 0.00000E+00 100 0 0.342 0.367 0.000 0.028 0.00000E+00 0.29587E-02 0.00000E+00 110.0 0.340 0.367 0.000 0.027 0.00000E+00 0.29433E-02 0.00000E+00 120.0 0.342 0.367 0.000 0.025 0.00000E+00 0 29361E-02 0.00000E+00 130.0 0.342 0.367 0.000 0.024 0.00000E+00 0 2932BE-02 0.00000E+00 140.0 0.342 0.367 0.000 0.023 0.00000E+00 0.29312E-02 0.00000E+00-150.0 0.342 0.347 0.000 0.021 0.00000E+00 0.29305E-02 0.00000E+00 160.0 0.342 0.347 0.000 0.020 0.00000E+00 0 29301E-02 0.00000E+00 170.0. 0.342 0.367 0.000 0.019 0.00000E+00 0 29299E-02 0.00000Ff00 180.0 0.342 0.367 0.000 0.018 0.00000Ef00 0 29299E-02 0.00000E+00 190.0 0.342 0.347 0.000 0.016 0.00000E+00 0.29298E-02 0.00000E+00 200.0 0.342 0.347 0.000 0.015 0.00000E+00 0 29298E-02 0.00000E+00 210.0 0.342 0.347 0.000 0.014 0.00000E+00 0 29298E-02 0.00000E+00 220.0 0.342 0.367 0.000 0.013 0.00000E+00 0.29298E-02 0.00000E+00 ACCllHilLATED FRACTION OF DEBRIS IN! Il DRYWELL(OllTSIDE CRD ROOM)= 0.00000E+00 WETWELL POOL = 0.10332E+01 1
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O O O O O O O O O O O Case 10. ENTER IPRESS,DT,TMStlHP,TMCOVR,TMSILL,NBRKS, RRRKO 0,,,,,,,,,,, ENTER INITIAL MOLTEN MASS IN RPU(KG)
. GRAVITY DISCHARGE CASE T(SEC) RBRK(M) ABRK(M*M) HRPV(M) - ,HCRD(M) GBRK(M**3/SEC) GDUNC OSILL 1.0 0.060 0.011 2.550 0.002 0.47776E-01 0.00000E+00 0.00000E+00 2.0 0.065 0.013 2.546 0.004 0.56917E-01 0.00000E+00 0.00000E+00 3.0 0.071 0.014 2.541 0.007 0.64834E-01 0.00000E+00 0.00000E+00 4.0 0.076 0.018 2.536 0.010 0.7751RE-01 0.00000E+00 0.00000F+00 5.0 0.082 0.021 2.530 0.014 0.RR956E-01 0.00000E+00 0.00000E+00 6.0 0.087 0.024 2.523 0.01R 0.10114E+00 0.00000E+00 0.00000E+00 7.0 0.093 0.027 2.516 -
0.023 0.11404E+00 0.00000E+00 0.00000E+00 8.0 0.098 0.030 2.507 0.028 0.12746E+00 0.00000E+00 0.00000E+00 9.0 0.104 0.034 2.497 0.034 0.14196E+00 0.00000E+00 0.00000E+00 10.0 0.109 0.037 2.487 0.041 0.,15493F+00 0.53209E-03 0.00000E+00 11.0 0.115 0.041 2.475 0.048 0.17255E+00 0.1543RE-01 0.00000E+00 ! 12.0 0.120 0.045 2.442 0.054 0.1RR78E+00 0.39896E-01 0.00000E+00 , i 13.0 0.125 0.049 2.448 0.060 0.20560E+00 0.69009E-01 0.00000E+00 l 14.0 0.131 0.054 2.432 0.066 0.22298E+00 0.99916E-01 0.00000E+00 l 15.0 0.136 0.058 2.416 is.071 0.240RRE+00 0.13085E+00 0.00000E+00 16.0 0.141 0.063 2.39R 0.075 0.25926F+00 0.16085E+00 0.00000E+00 17.0 0.147 0.048 2.379 0.079 0.27809E+00 0.18950E+00 0.00000E+00 18.0 0.152 0.073 2.358 0.083 0.29732E+00 0.21675E+00 0.00000E+00 19.0 0.157 0.078 2.336 0.086 0.31691E+00 0.24273E+00 0.00000E+00 4 20.0 0.163 0.083 2.312 0.089 0.336R1,E+00 0 24766E+00 0.00000E+00 21.0 0.168 0.089 2.287 0. 092- 0.35696E+ 00 0 29175E+00 0.00000E+00 22.0 0.173 0.094 2.261 0.094 0.37731E+00 0.31521E+00 0.00000E+00 23.0 0.17R 0.100 2.233 0.097 0.397R1E+00 0.33821F+00 0.00000F+00 24.0 0.183 0.104 2.203 0.100 0.4183RE+00 0.340RSE+00 0.00000E+00
O O O O O O O O O O O l l l l-5 Y [gb SU O le!Nf f/LL 25.0 0.189 0.112 2.172 0.102 0.43R96E+00 0.38323E+00 0.00000E+00 26.0 0.194 0.11R 2.139 0.104 0.45949E+00.0.4053RE+00 0.00000E+00 27.0 0.199 0 124 2.104 0.107 0.4798RE+00 0.42732E+00 0.00000E+00 28.0 . 0.204 0.130 2.068 0.109 0.5000RF+00 0.44903E+00 0.00000E+00 r 29.0 0.209 0.137 2.029 0.111 0.51998E+00 0.47049E+00 0.00000E+00 l 30.0 0.213 0~143
. 1.989 0.113 0.53951F+00 0.49163E+00 0.00000E+00 40.0 0.259 0.211. 1.4R6 0 129 0.69163F+00 0.66790E+00 0.00000E+00 50.0 0.297 0.?76 0.73R 0.128 0.65022E+00 0.6857RE+00 0.00000E+00 60.0 0.310 0.302 0.000 0.059 0.00000E+00 0.1227RE+00 0.00000E+00 70.0 0.310' O.302 0.000 0.03R 0.00000F.+00 0.19889E-01 0.00000E+00 00.0 0.310 0.302 0.000 0.033 0.00000E+00 0.78971E-02 0.00000E+00 90.0 0.310 0.302 0.000 0.030 0.00000E+00 0.48182E-02 0.00000E+00 l 100.0 0.310 0.302 0.000 0.02R 0.00000E+00 0 3733RE-02 0.00000E+00 110.0 0.310 0.302 0.000 0.027 0.00000E+00 0.32904E-02 0.00000E+00 120.0 0.310 0.302 0.000- 0.025 0.00000E+00 0.30955E-02 0.00000E+00 130.0 0.310 0.302 0.000 0.024 0.00000E+00 0.30068E-07 0.00000E+00 140.0 0.310 0.302 0.000 0.023 0.00000E+00 0.2965RE-02 0.00000E+00 150.0 0.310 0.302 0.000 0.022 0.00000E+00 0.29466E-02 0.00000E+00 160.0 0.310 0.302 0.000 0.020 0.00000E+00 0.29377E-02 0.00000E+00 170.0 0.310 0.302 0.000 0.019 0.00000E+00 0.29335E-02'O.00000E+00 180.0 0.310 0.302 0.000 0.01R 0.00000E+00 0.79315E-02 0.00000E+00 j
190.0 0.'310 0.302 0.000 0.016 0.00000E+00 0.29306E-02 0.00000E+00 200.0 0.310 0.302 0.000 0.015 0.00000E+00 0.29302E-02 0.00000E+00 210.0 0.310 0.302- 0.000 *0.014 0.00000E+00 0.29300E-02 0.00000E+00 220.0 0.310 0.302 0.000- 0.013 0.00000E+00 0.29299E-02 0.00000E+00 ACCUMULATED FRACTION OF DEBRIS IN: DRYWELL(OUTSIDE CRD ROOM)= 0.00000E+00 WETWELL POOL = 0.81493Ef00
)
? . i i ! l > l l l l APPENDIX C ? 9 1
O
! APPENDIX C O A SELECTED REVIEW OF THE PROBABILISTIC RISE ASSESS!,1ENT PREPARED FOR THE SHOREHAIs liUCLEAR P0i(ER STATION O
- O
- Prepared for Future Resources Associates, Inc. ,
Berkeley, California-
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- O t
i by P. R. Davis .O l l O August 1982 4 !O 1 P i
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+--w-9--* y -ww _.e-+ 4 ..__m4 ._.,,,.,,_,-,_e ew- - - ,y -.-&y --mm -t- -.~,-
O 10 T A37.E _03' CONT:'NTS
'O I. IliTRODUCTION . . . . . . . . . . . . . . . . . '. 1 II. C01.IPARI30N iiITii OTHER PRA RESULTS . . . . . . . A g III. PRSLI!!Ili ARY REVIEli - QUESTI0li3 AND ANS,iERS . . . 6 IV . AD7ITI0HAL ISSU:'S COHSIDERO Ili TE.' REVIEV1 . . . 18 ;g V. 00liCIUSIOI.~5 . . . . . . . . . . . . . . . . . . 22 O
I i l
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i { 1 \ ? \O i l i k l l l I,
O O 3 A SELECTED REVIE# OF THE
.LROE431LISTIC ISL ASi'Si ENT F1EPARED FOR Tg SHORSHAD liUCL3AR FO.VER STATION
- 1. INTRODUCTION O
This report presents the results of an effort to parfarm a limited review of selected areas of the shoreham Nuclear g Fower Station Probabilistic Risk Assessment (FRA). The effort wa, undertaken in June 1982 based on a servicer screement between the author and Future Rercurcer Acsociaten (FRA). FRA 3 h?.d previously entered into a contract with suffolk County, New York (the site of tha shoreham station), and this effort is part of a broadar task involving other contractors and g conrultants to fulfill the Suffolk county contract. ~ Due to time 11.ritations, only ten man-days were available for the review and report preparation. This is innufficient 3 to provide a comprehensive review of all important areas of the FRA. However, based on the authors experience,1 it was cone]udad that a meaningful, though limited, review could be D l The author has performed some nine reviews of Frobabi-listic Risk Assessments for various clients. Includ'i in this nine have been four reviews of B.VR FR As similar to Shoreham. In addition, the author has performed core melt probability 3 assessnents of two EnRs. D
- O 2
. accomplished. The review was linited primarily to a consider-ation of core melt probability and initial containment response. .O The review proceeded basically in three phases. The first phase consisted of comparing the core damage probability estimated by the Shoreham.PRA with similar estimates from four .O other PRA studies performed for boiling water reactors. The purpose of this comparison was to determine if the Shoreham PRA result was consistent with similar studies involving different ;O sponsoring organizations and contractors.
The second phase involved developing a list of questions, based on a preliminary review, which was transmitted to the O PRA performing contractor (Science Applications, Inc.) and a response obtained. The third and final phase consisted of identifying any O remaining issues and considerations which were judged to have the potential for influencin6 either the probability or consequences of core damage accidents as assessed in the O Shoreham PRA. The report is organized around these three phases, with l Sections two through four covering the phases. A final section, lO l Section five, provides the co:.clusions of the review. It should be noted at the outset that performing probabi-listic risk assessments for nuclear power plants involves a lO considerable amount of subjective jud 6ement and interpretation. While significant advancements have been made in both method-ology and data processing, the PRA field is of relatively recent ,0 origin, and a substantial body of ongoing research is providing l lO
E O 3 new insights on various issues. As a result, there is a considerable difference of opinion among PRA practitioners regarding the validity and uncertainty of various FRA methods. In many cases, these differences can be shown to have a negligible effect on risk. In any event, most scientists and engineers familiar with FRA can find areas of disagreement and criticism with any FRA study. This does not necessarily mean the results are invalid, it merely reflects the subjective nature of many PRA methods and procedures. 40
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G[ 4 4 , V U U. U U U O O 4 Table II-l - PROBABILISTIC RISK ASSESSMENTS CURREHTLY AVAILABLE - NSS Plant Comercial Power rendor Utility Type A-E Operation Level Publication Sponsor Performer Remarks Browns Ferry 1 GE TVA BWR4/ Mark I Utility 1973 1065 Aug. 1982 NRC Grand Gulf 1 EG&G IREP Plant GE Mississippi BWRS/ Mark 3 Bechtel 1982 1250 Oct. 1981 Power & NRC Sandia RSSMAP Plant L19ht Peach Bottom 2 GE Philadel- BWR4/ Mark I Bechtel 1973 phia Elec. 1055 Oct.1975 NRC NRC-var- RSS plant ious cont. Limerick GE Philadel- BWR4/ Mark 2 Bechtel 1985 1065 1982 Utility phia Elec. SAI/PLG - Shoreham GE LILCO BWR4/ Mark 2 S&W 1983 819 1982
- Utility SAI/PLG Draft only GESSAR II available ! GE (NA) BWR6/ Mark 3 (NA) (NA) 1220 March 1982 GE e GE Contains proprietary ' information Sequoyah W TVA PWR-4 loop Utility 1982 1148 Feb.1981 NRC Sandia RSSMAP Plant (IceCond.)
Indian Point 2 W Consoli- PWR-4 loop UE&C 7/74 873 1982 Utility PLG dated Ed. Indian Point 3 W PASNY PWR-4 loop UE&C 8/76 965 1982 Utility PLG Zion 1 & 2 W Comon- PWR-4 loop S&L 1973 1040 1981 Utili ty PLG wealth Ed. Surry W Va. Elect. PWR-3 loop S&W 12/72 775 Oct. 1975 NRC NRC-var- RSS Plant
& Power ious cont.
Crystal River 3 B&W Florida PWR Gilbert 3/77 825 Dec. 1981 NRC SAI IREP Plant Power Oconee 3 B&W Duke Pc,wer PWR Utility / 12/74 886 Jan. 1981 NRC Sandia RSSMAP Plant Bechtel 4 e e l
4A Table II CDU.FARISDN 0? CCRE D'J: AGE PR03A3ILITIES FOR BOILIHG VIATER R.: ACTORS D Plant Type FRA Sponsor Principal Damage C'ntrg,ctor Browns Ferry BWR 4/ Mark I Nuclear Regulatory EGLG 2.0 X 10~4 Commission D Grand Gulf 3?iR 6/r. ark III URC Sandia 3. 7 X 10 -5 Labs Peach Botton BdR 4/Eark I URC (1) 2.9 X 10 -5 '~) Limerick B7/R 4/Eark II Ehiladelphia SAI 1.5 X 10-5 Electric Co. Shoreham Blia 4/ Mark II Long Inland SAI 4.4 X 10 -5(2) Lichtin2_Co. s so-t' l Various contractors involved, under direct NRC canagement 'J 2 This value is actually a " core vulnerable" condition, as described in the Shoreham PRA. A small factor is applied to each core vulnerable probability to estimate the core damace probability. None of the ot'r ' PRAs listed used this additional factor. Thus, the Shoreham "corc irerable" probability is comparable to the other values in the table for core dama6e. (See Section IV for further discussion of this approach.) l O
O 5 II. COMPARISON ,*4ITH OTHER PRA RESULTS This section presents the results of the first phase of the Shorehan PRA O review; a comparison of the Shoreham core damage probability estimate with that computed by other studies. To date some 13 PRA studies have been com-pleted and published. These studies are shown in Table 11-1 along with per-tinent information relative to the plant design. As shown in Table II-1, a o total of six BWR PRAs have been published. However, the GESSAR II PRA contains proprietary infomation, and will not be considered further in this comparison. The remaining five BWR PRAs are compared in Table II-2. The second column (Type) indicates firts the primary system design classification (BWR 4 or BWR 6) o and the containment design (Mark I, liark II, or Mark III). For purposes of core damage probability estimates, the primary system design, along with ass-ociated safety systems, is more important than containment design. The third column indicates the organization which sponsored (funded) tht o PRA effort, and the fourth column indicates the principal contractor who per-formed the work. The last column is the overall result in terms of computed probability of core damage accidents per reactor year for the plant being considered. As indicated by the last column in Table II-2, the Shoreham re-O sult falls in the middle of the core damage probability distribution, being about a factor of three above the lowest (Limerick) and about a factor of three below the highest (Browns Ferry). It should be noted that the Shoreham PRA uses an additional factor O which reduces somewhat the Table II-2 computed probability. This factor . is applied to the " core vulnerable" condition, which is the value in the result column of Table II-2. None of the other PRAs use this approach. The use of and validity of this factor is explored further in Section IV. lO !O O 'O
e 6 III. FRSLI!.iIRARY RSVIEd-QUGTIOi 5 AND AH3 JARS
, Following a preliminary review, a list of seven questions was formulated which were transmitted to the Shoreham PRA contractor (Science Applications, Inc.). It should be noted , that many more than seven questions were derived from this review, but all but seven were eliminated based on their resolution as a result of other information found in the PR4, , or were found to not naza a significant potential influence on either the probability or consequences of core dana ~e.
(Further potentially important issues were identified as the
"*"i'* P" **d*d' Th*** ""* Oi" "" *d I" th' # ll "i"' ** *i""'
O l Insufficient ti.r.e was available to discuss and attempt to resolve these issues v.ith the contractor. ilowever, the O " l'II"6 " *i " f *'i "*F "* ****P* *" 1" * *h" potential impact of these issues.) The seven questions formulated in the preliminary raview g are considered separately. A brief explanation of the poten-tia] sicnificance follows each question, followed by the SAI response, and concluding with a discussion of the adequacy g of tne response.
- 1. A.- ;uestion -#ill the hi 6 h drywell temperatures calcu-lated to occur during degraced core cooling accidents cause g rapid degradation of the concrete wal] and a pressure buildup inside the drywell?
1.B.-Explanation-Concrete is not a high temperature g structural material. De grada tion, in terms of hydr,. ding and C
P 7 i decomposition, can start as low as 200* F depending on age and composition of the concrete. While drywell temparatures are ? not displayed in Appendix C (MARCH calculation results), a , drywell temperature of 3000* F is indicated on page C-73 (although this value is said to be unrealistic). 1.0.-SAI Response-No. Upon concrete decradation due to high temperature, H2 O is released from the concrete between 200'F and 700*F, and 00 2 is released between 1200* F and 1700* F. In onder to increase the co.rtainment pressure by 1 pai, conser-vatively assuming no heat loss through the pass'ive heat sinks and no simtion in the concrete, approxitately 105 cubic feet. of concrete at an elevated temperature of 200-700*F would have to release its free water. At more elevated tenperatures (creater than 1200* F) the deconposition of approximatel: 60 cubic feet of concrete would increase containment pressure about 1 psi. In terms of the unlined concrete walls, the thermal bondinc layer above 1200*F should be approximately 6 inches deep for pressures to exceed the containment ultimate ' pressure capacity during periods of hi 6h containment temperaturcs. The particultr sequence for which high temperatures are significant before vessel heat failure is Class 3. The thermal transient duration for this secuence, however, is very short. The re fore , the concrete wall is not expected to reach the decomposition temperatures of the carbonates to a depth of 1/2 foot. 1.I).-3 valuation of Response-The SAI responce seems adecuate. However, since drywell temperatures and the resulting concrete thermal response is not provided in the report, it is D
8 not possible to confirm that the rotential for high tenperature drywell degradation is insignificant. A further examination of this issue (see Section IV), tends to indicate th't it may not be of concern.
- 2. A.-juestion-How much effect on suppression pool decon-tacinatior actors does the normal IJARCH code calculation of very litti heat transfer between the hot noncondensible gases
~
have ? 2.3.-Discussion-Generally, the EARCH code calculates very little heat transfer between toe hot non-condensible gases and the supuression pool throu,h which the gases would flow folJowing a de rraded core cooling accident. This can produce unrealistically Jow sup;ression pool tempernturec. The decontauination factor (DF) for radionuclides in the suppression pool is thoucht to be decraded as the suppreasion pool temperature approaches naturction conditions. 1;eglecting gas to pool heat transfer may therefore result in acre optiaistic DFs then should realis-tically be expec ted . 2.C.-3 AI dSSFO:!SE '2he sensible energ/ of the hot noncon- , densible cases flowing from the pricary system or drfwell into the wetwell throu;h the S/RV or vents, respectively, are considered in the containment responne analynis of Shorehan. These gases are conservatively assu. sed to leave the suppression pool at suppression pool temperature, effectively cooling th.7 gases down. The suppression pool temperature correspondingly increases due to this ener,x addition. Che decontamination factor that was used accounted for this additional heat up of tr.e wetwell pool due to the noncondensible gases.
O 9 2.D.-Evaluation of Response-The SAI response appears O adequate and the issue is considered resolved. Furthermore, the assured SAI DFs (Table 3. 7. 4, pp. 3-160) seen quite pessicistic (too low) based on recent information. This issue is discussed O further in section IV.
- 3. A.-Question-Have high temperature penetration failures and pre-existing containment leaks been considered?
O 3.3.-Discussion-Relatively recent evaluations have shown a significant number of instances when reactor containment building penetrations have been found leaking. Further, O penetration seals are not designed to withstand the severe degraded core accidents analyzed in the FRA. 3.C.-SAI aceponse-plant response and potential contain: cent I) failures due to hich teaperature degradation of penetration seals were not specifically analyzed for plant consequences. These were considered implicitly in the assigned probability O of containment leaxage which precludes overpressurization as
- snown in the containment event trees.
3.D.-Evaluation of Responqe-The SAI response is not O considered adequate to resolve this question. However, it l appears that direct pool scrubbing of most radionuclides (by release through safety and relief valves) will occur for the O most probable accident sequences (Table 1, pg. 3). In these cases, containment leakage is of lesser consequence. Furthernore, the Reactor Buildin6 Standby Ventilation System can procens C) and filter significant amounts of containment leaka6e (see following question), especially if pool scrubbing has previously O
O 10 occurred to reisove aerosols and fission products.. g 4. A.-Question-#ith dedradation of the Reactor Building Standby Ventilation System (R33VS) from aerosol loadinc, high-fission product loading, and' adverse environnent, can the O syste keep up with modest leakage ? 4.B.-Discussion-The RBSYS is designed to control the minor radionuclide releases which can result from normal oper-O ation, in addition to those releases predicted to occur from design basis accidents. It is not designed, however, to operate under degraded core cooling accidents when excessive pressures, 9 te mpe rature s , and aeroso3/radionunlide filter loadings may challenge the systen far beyond its design cepabilities. 4.C.-Response-The RESVS ventilation system consists of a three parallel 45000 cfn exhaust fans (e,ch havir r 2003 capacity wnen used for recirculation during abnormal conditions) and two 1585 cfm 100) capacity parallel filter trains. r,ince g each fan is 100) capacity for the R33VS mode, only one filter train fan is re. quired to operate, and the redundant filter trains may be shut down. The potential filtration of fission g products was considered in the PRA tr the extent that t've R3SYS cculd handle design leakage ratas and partial contain. ent failure leaka.;c rates which do not overpressurize the reactor - g building. At these flow rates, the dilution factors of these 6ases containini- fission products and aerosole are on the order of 1500 to 1. Additionally, for these accidant scenarios, it is known that aerosol pluising of tne leakage paths can ocaur, () which is expected to reduce the fisaion product and aerosol O
O-11 concentration of -the gases escapins to the secondary enclosure. O Therefore, failure of the R33Vs was considered a low probabf'.ity event (i.e. 0.01 for C 1 and C) accident sequence classes).
~
4 However, degradation of the R33VS performance was addressed in e x3 the ana3ysis of reduced decontaminntion factors for filter efficiencies. 4.D.-3 valuation of Response-The SAI response is considered C marginally adequate to resolve the issue. The PRA assessment of RESVS design capability and potentiel failure moder is in-sufficient to determine if a rigorous evaluation of RBSYS O availability has been made. However, considerinc that the most probable accident secuences appear to be those which dis-charge most radionuclides directly to the supprercion pool, O and that the suppression pool DFs selected by SAI appear to be pessimistic, this issue seems to have little potential impact on risk. O 5. A.-Question-Appendix H of the Shoreham FRA assumes that the hot debris bed which can be formed from core materials is always coolable if sufficient water can be supplied. What
- <y v is the basis for this assumption, and has the effect of enhanced radionuclide release with attendant steam pressure excursion on containment integrity when this coolinc occurs been considered?
C) 5.B.-Discussion-During the progression of a core meltdown i accident, it is litely that following penetration of the reactor vessel, the core material may form a debris bad which $3 could be coolable providing water can be supplied . ,0
J 12 5.C.-3 AI Response-The .ippendix H essumption of debris bed coolability, niven that a sufficient water supply is avai]able, is based upon previous analyses of coolable debris bed formation (references provided) and engineerin6 judgment. Uhese analyses indicate that the dryout heat flux of a debris bed is a function of tne bed height, the porosity, and the averace particle sizr: of the debris. II.edium-sized particles resulting from vessel head failure or from mild core-water interaction may fora a coolable debric bed if the decay heat level is low, and if the debris bed is not isolated from the water. Chc bubb'ing of gases, which are generated fra: concrete attack, throui;h the celt could agitate the molten core catorial allowin , water to penetrate an6 requench the Delt jnto coolable rarticles.
~
D The actual ,:eouetric configuration of the shorehan con-tainaent, specifically the drywell floor and the pedestal reginn, was considered in characterizing the debris bed foraation, the depth of the debris bec, and the probable cediu..-sized pr.rticles that can fora from a mild interaction of the core material with the water. D Analysis of the disposition of the core debris upon bottom head failure indicates that tne molten core debris would flow out of the pedestal region. The debrin could scread out into the drywell floor, and a portion could also flow into the wet-well pool through tne do-ncomers located unterneath the R'Y in the -nedestal recion. The hei:;ht cf tha core nat'ria' that can D accu.iulate on the drywell floor is limited by tha downcomer confi ,uration to a maximum height of six inches. D
5 j 8 13 The downcomer lip which extends about 6 inches above the floor could nelt due to the interaction with the high temperature 9 core debris. If this occurs, the debris bed heicht would be determined by the height of the flange which is on3y about 2i inches. There fo re , it was concluded that a very shallow S bed would form, assumin6 long term coolin(; to the water over-burdern, given sufficient coolant injection. 5.D.-Evaluation of Response-The S AI respense to the debris G bed coolability issue is reasonable, and valid argunents are given supporting the potential for coolability. However, since the geometry and particle size of the debris bed is D unsnown, assessments of coolability must be considered specu-lative. SAI did not respond to the part of the question dealing with steam pressure spikes and fission product release. S Upon further consideration of this issue, it la considered not to be significant in terms of the potential for risk increase for the following reasons: J 1. If containaent failure nas not occurred at the time debris bed cooling occurs, most of the fission products (which are released during the initial meltdown phase) will be securely trapped in the suppression pool water. Thus, the possibility of containment failure froa a steam pressure surge upon debris bed coolin o would not cause a large fission product ) release.
- 2. If containment failure occurs before debris bed cooling, the major consequer ces of the accident would be underway, and D
the added fission product release would lixely not be significant. )
O 14
- 3. A steam pressure surge sufficient to challenge con-O tainment integrity requires a large amount of water delivered to the bed and intimate mixing. It is not likely that a hi Gh volume water source would be available since most such sources O within the containment must previously have been assuned failed or degraded to cause the accident to progress through core meltdown.
LO 6. A.-ouestion-Have the recent Sandia experiments showing the potential for energetic molten core-water reactions been considered in the Shoreham PRA? Also, what is the basis for O the particle size assumption of in (diameter) in the vessel and 2" in the reactor cavity? 6.B.-Discussion-This question is related to the previous O one, except that tne concern here is that energetic molten core water reactions may occur within the vessel or upon vessel failure if water already exists at tnese location. These events 'O could cause early containcent failure if the steam pressure surges were sufficient. 6.0.-S AI Response-The Sandia experiments on molten core-O water interaction were considered in the Shoreham PRA. For very energetic molten core-water interaction, the mode of melt delivery into a pool of water and mixing with the water can O affect the break up and interaction of the fuel and wnter. It is also known that the initial conditions can suppress efficient core-water fragmentation. In the Shoreham PRA, this in-vessel
- () interaction was considered probabilistically. The source term and release characteristics for the base cese were extrapolated from WASH-1400 anal
- ' sis. However, the specific accident O
O 15 sequence and the conditions in the vesse) at the tiue of potential core-water interaction were considered in assicning O the conditional probability. That is, for a high prersure transient accident.sequt,ce, the probability of in-vessel steam explocion was considered lower than that for a large LOCA I) sequence in which the primary syeten pressure is at appraximately the containment pressure. The containment response was analyzed for an ex-vescel O care-water interaction in the drywell. Ex-vensel fuel-coolant mixing was considered more likely if water was available in the drywell. The analysis of the efficiency of the fracaentation, O however, considered the geonetric configuration of the floor. The shallow water pool could inhibit a very efficient mixing of coolant with the molten core. It was assessed thnt this O would retard the rapid and efficient fuel fracmentation that is needed to form more heat transfer area for explosion propa-getien. An overall mean particle size of 2 inches was judgad O to approximate the particle size distribution for rapid steam j formation during this period . Additionally, a sensitivity analysis was performed to assess the effect of small particle IO size formation. Such formation resulted in containment failure inaediately following vessel heat failure. This sequence was accounted for in the containment event tree and included in O release category AA. 6.D.-Evaluation of Response-The SAI response seems some-what confusing, and is considered only partially adequate. O However, a further evaluation of this issue indicates that the potential for large energetic reactions is small. Since 10 l .- .
P O 16 extensive and complex structures exist in a 3'.7R below the core, O it is unlikely that the molten core migration could proceed in a coherent manner. LarGe, energetic molten core-water reactions require that rather large coherent masses of molten material O and water mix in a short time. Furthermore, these below core structures contain a significant amount of stored heat and also provide extensive heat conduction paths from the core O region to the lower plenum. It is therefore likely that most or all of -the lower plenum water would be boiled off before the arrival of significant amounts of molten material.
- O In the reactor cavity below the vessel, vertical open pipes exist, connected to the suppression pool. Thus , it is not likely that any si dnificant amounts of water could exist ID in this re6 1 on to interact with molten material upon vessel failure (which in not lixely to be coherent due to the many penetrations in the lower head) .
'O 7. A.-Question-iVhat is the basis for assu.aing (pg. c-23) no sircaloy-water reaction wnen the zircaloy becomes molten?
7.3.-Discussion-The reaction between zircalon and water c0 can be important because it increases the heat Generation rate in the core and produces hydrogen. 'ihe resulting hydrogen can l l burn, adding additional heat load to the containment, or it can 'O decrease the D.F. of the suppression pool if it mixes and flows with radionuclides to the pool. 7.0.-S AI Response-The statement on p. C-23 is as follows: X3 "idetal-water reaction in a melted node was not allowed." During the postulated core degradation and neltdown accident i
?O
5 lO ! 17 l. , scenario, the core. heat up and meltdown model assumes a forma-1. !c) ~ tion of a coherent molten fuel pool supi.orted by a crust at the solid fuel-melt interface. This crust and molten layer ! causes the steam blockage which prevents water or steam from d)- flowing throu :h the channels with molten fuel . . Therefore, i i metal-water. reaction in a melted node (or ZR) was not allowed. ! 7.D.-Response Evaluation-The SAI response is considered O adequate in resolving this question. O O O O O P O O
O 18 IV. ADDITIONAL ISSUES CONSIDERED IN THE REVIE7I 'O l t This section considers additional issues which were identified as potentially important during the review and l assesses their potential si Cnificance. These issues have
)
not been discussed with the FRA contractor since time. was t not available. The issues are as follows: l 1. ESIV Leakage - Durin,: essentially all accidents considered in the Shorehan paA, the 'J.nin Stean Isolatian l Vc1ves are assu ned to close based on the 1;SIV closure logic incorporated into the plant design. LSIV closure isoletes the primary systen from the remainder of the power conversion l system (turbine, condensor, feedwater pumps, e tc . ) . For ntny of the higher probabi'ity accidents, the effluent fron n the prinary system (w'aich contains racionuclides ralansed from the overheated core) is relecsed to tha suppression pool throu;h the safety and relief valvec. If the LSIV seal is m-"
. effective for these accidents, nost of the fission products are scrubbed out by the po oi'. However, pas t experience with M3IVs has shown that lear:a.:e occasionally occurs. If such a
3 eakage occurred during the accidents being conciderad, a potential path through ti;e valves and to the at.osphere cou?d exist. Tais path could be even nore likely when it is O recognized that the :l.SIVs are not designed for the high teL.,erature radiation environnent produced by the eccidents considered. The Shorehan FRA does not appear to irovide en O adequate consideration of this issue. O
O 19 Two questions are involved in resolving this issue: What (3 is tae likelihood of si ;nificanti LSIV leakage during important f accidents, and what is the consequence of such leakage in terns of off-site release . O Due to time constraints, it was not possible to explore either of these questions in depth. However, it has been recognized that 17.SIV leakage does occassionally occur, and O desi,;n and uaintenance inprovements are underway to provide additional assurance of leak intecrity. preau mbly, the Shorehan Plant will take advantage of these improve.ients. !O With resrect to the second question, II.SIV leakace does not autocatically mean release to the atuonphere. In fact, the systea downntream of the ESIV is bEsical]y a closed system I O during operation, cnd it is desi Sned to preclude radioectivity ralease. -lhile potential paths to the atmosphere may exist under some conditions (such as condensor air ejectors) such
- O leakage is expected to be automatica1?y isolatad and can a? so be manually isolated.
The potential for leal:a:e through the IT.SIVs to the i 'O atmosphere depends on system deci;n and operating procedures. However, based on this very perfunctory consideration of the i
- ar.tter, it does not appear to be a significant potential
() contributor to risk. [
- 2. Control system Pailure - All 3WRs have rather l
l complex control systems which monitor the flow of energy fron O the core to the turbine, the core power, the feedwater flow, l i O
,n- n ~
20 and various other parame ters. The systen automatically adjusts some of these parameters to compensate for changing conditions and to keep the plant running smoothly. It has been recognized rather recently that this same system can malfunction and cause i the ^ulant to experience a serious transient which could lead ? to a degraded core condition. This potential event does not appear to be adequately considered in the Shorehan pRA, or any other PRA. ilhile it has not been conclusively demonstrated that such events have the potentin1 for causing severe core damace which would significantly contribute to risk, some control system malfunctions have occurred which have caused rather severe disturbances to the normal operation of the plant. Until more work is completed (currently underway within URC and its contractors), it is not possiDie to determine the significance of this potential for shorehau.
- 3. Dependent Failures - One particularly difficult issue with all pHAs has been the treatment of dependent failure, wherein the failure of one system causes or influencas the failure of another. The shoreham PRA is considered deficient with respect to the description of how dependent failures were identified and quantified. It is not possible to tell, based on the information provided, if an adequate assesstent has been given to dependent failures.
- 4. Decontamination Factors - The Shoreham FRA assumes suppression pool decontamination factors ranging from 1 to 100 (pp. 3-160). Recent research in this area, including experiments supported by analysis, shows that DFs two or more
)
O 21 orders of magnitude higher would likely be more realistic. O '#hile such increases have not been universally accepted, the potential for significant increase appears strong. -This would markedly decrease the source term for many accident sequences. <J 5. Derivation of " conditional Frequency of Release" - The Shoreham FRA assigns (Table 4.2) values ranging from .07 to 1.0 to accident sequences. These values are multiplied by the O calculated probability of a core vulnerable condition to arrive at the actual probability of a given accident sequence. The factor (pp. 3-142, 3-177) is apparently to account for the O possibility that the core may not melt under the conditions assumed, or by some undefined mechanism, sufficient coolant is re-introduced into the system to terminate core heat-up. f) Jhile this is a novel approach which appears to have merit, an adequate description of how the " Conditional Frequency of Release" values were derived could not be found. The factors ,O do not appear, however, to have an overly significant influence on computed risk since they are not very small.
- 6. Conflict with Recent Jegraded Core Probabilistic
~O Assessment - Recently, an URC sponsored study (EUREo/CR-2497) has suggested that core damage probabilities may be significantly higher than assessed in the shoreham (and other) FRA". #hile O this report is under review (the author is involved in one such review), preliminary indications are that the results are based on invalid extrapolations, include questionable statistical O procedures, and do not consider systens and procedures which can be effective in mitigating the progression of core meltdown accidents. .O
)
22 V. 001;CLUSIONS
)
Based .cn2 this limited and selective review of the Shoreham PRA, the following conclusions are derived:
- The study appears to be a comprehensive and competent
)
assessment eaployin; state of the art nethodology and generally adequate data sources.
- The results were found to be comparable, in terms of
) dcCraded core condition probability, to other PRA studies of similar plants.
- dhile some areas of the study were considered to be
) deficient, in no case was a deficiency found which appeared to have a sirnificant potential for inerensing tne risk. (The omission of external event accident
) initiators could be an exception, but review of thic issue was beyond the score of the effort.)
- One area (suppression pool decontamination factors)
) was found which appears to have a significant rotential for decreasing the computed risks from Shorehan.
.4ecent evidence strong 3y suggests thet nuch hicher
) DFs may be appropriate for suppression pools. This could substantially reduce the shoreham radionuclide release. ) ) )
D D D Appendix 0 INTERNAL FLOOD ANALYSIS S by HOWARD E. LAMBERT D D D D D D D
- O t
!O Appendix D i Table of Contents 1 !O i I . i D.1 Introduction j D.2 Description of the Internal Flooding Sequence D.3 Units of Frequency Issue (O l D.4 Level of Stress Issue D.5 Observations
- D.6 Alternate Probability Expression
!O ; I !O 4 .,0 i
- O '
I l lO I { l 1 f
- O t
1 i lO 4 4
.O APPENDIX D g INTERNAL FLOODING ANALYSIS-(SAI's Section 3.4.4.1 and Appendix G)
- g D.1 Introduction This appendix addresses the possibility that portions of emergency core cool-ing systems could be disassembled during maintenance (e.g. a pump impeller re-O placement, valve stem replacement, valve seat adjustments). If during' this dis-assembly, human error or set of human errors occur which deisolate the component undergoing maintenance, such as opening a M0V, then release of water through the opened valve can occur. This has already occurred at Peachbottom.
10 . As described below, we believe that SAI's computation of the accident frequcncy caused by internal flooding is incorrect and non-conservative for the two reasons described below.
- O e The approach used by SAI in quantification of these sequences does not produce units with accident frequency. One key event in the flooding sequence is the operator inadvertantly-
- O pening a valve during maintenance. The probability est-imate for this event is given on a per maintenance act basis.
SAI's calculations do not reflect this, which leads to a factor of about 100 in underestimating the internal flood frequency. .O e SAI used human error probability estimates from Swain and Guttmann's work that do not reflect a highly stressful sit-uation. We believe that a highly stressful situation does
- O exist when internal flooding occurs and that all calculations should reflect this.
O D-1
~O
O D.2 Description of the Internal Flooding Sequence lO SAI identified 8 initiator types which would lead to flooding at level 8 in the Reactor Building where several ECCS and residual heat removal components are located. In particular, SAI identified 3 initiator' types, 2, 3 and 4 as described below which could quickly (within 30 to 40 Minutes) flood and disable the compon-O ents described in SAI's Table G.6 if the operatur fails to reclose the isolation valve in time. INITIATOR WATER SOURCE SYSTEMS INVOLVED * .O 2 Containment Storage Tank HPCI, RCIC and CS 3&4 Screenwell RBCLCW & RHR O In this case, it should be noted that internal flooding does not result in loss of the power conversion system, a high pressure injection system normally on-line when the reactor is producing power. While flooding occurs, if the operator erroneously isolates the power conversion system (contrary to normal operating
- O procedures), then an accident sequence is initiated as described by the bold line on the event tree in figure 3.4.23.
'O \o O
- HPCI: High Pressure Coolant Injection RBCLCW: Reactor Building Closed RCIC: Reactor Core Isolation Cooling L op Cooling Water CS: Containment Spray RHR: Residual Heat Removal O
D-2 O
- r-,vr n ,e - , - , - - , , - - - - - - - - , , r - - , , -
b D e Table G.6
SUMMARY
OF VITAL EQUIPMENT ASSOCIATED WITH SAFETY SYSTEMS LOCATED IN THE ELEVATION 8 COMPARTMENT AND THE POSTULATED HEIGHT AT WHICH VITAL EQUIPMENT COULD DISABLE THE SAFETY SYSTEM e A550CIAT(D POSTULATED SYSTEM VITAL [Qu!PMENT D!$ASLED MEIGHT , MPCI HPCI PUMP 5 6'-0* HPCI Tutt!NE s' HPCI INST. RACK A.B *3'-0" RCIC RCIC PUMP 5'-0* RCIC TURBlNE 4'-0" RCIC INST. RACK A.B *$'-0* g LK! RMR PtePS 6'-0* RMR INST. RACK A.B *6'-0* CORE CORE SPRAY P' IMP 4' 0* SPRAY CORI SFRAY INST. RACK A.I '6' g
- Based on the assumption that a short util cause a shutsown cf the system.
** The estimated figures indicate that the suppression pool inventory may or may not be enough to effect vital equipment. Momever, if this inventory is costined with other sources such as the reactor primary system (during ADS sequence) tt util without question disable the system.
9 8 D D-3 e
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- ) of the power conversion system.
e Event A: On-line maintenance of any of the following systems occurs D 1 HPCI 2 RCIC 3 Core Spray 4 RBCLCW HX 5 RHR HX D e Event B: System is disassembled for maintenance, o Event C: Operator inadvertantly opens an isolation valve durint maintenance causing flooding to start. O e Event D: Operator fails to reclose the isolation valve within 40 minutes which results in flooding to the six foot level. e Event E: Operator erroqecasly isolates power conversion system during floodirj. D A typical fault tree development for the sequence cescribed above is shown in figure G.6. It must be noted that SAI combined events A and B on the fault tree. D D.3 Units of Frequency Issue As described in the previous section, event C is the initiating event which I causes flooding. The occurrence of events A and B defines a vulnerable system state that permits event C to initiate flooding when event C occurs. Further-more, the occurrence of events A and B and C defines another vulnerable system state that permits event E to initiate the accident sequence when event E occurs. Because event C is an initiating event, we must compute the frequency of occurrence of this event. D D-5 9 l l
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g Events C and D describe operator error and SAI used Swain and Guttmann's a estimates, reference 1, to assign probabilities to these events. Specifically for event C, SAI on page G-38 of the draft Shoreham PRA said "The conditional probability that the operator opens the isolation, valve to the system while it is undergoing maintenance is determined from Swain and Guttmann. The value is
- g determined to 5x10-3/ valve operation. This error rate is higher than normal plant operator error rates since it has been shown that operator error under maintenance conditions may be higher than that found perfonning normal plant function".
O Using the above probability, we must compute the number of valve operations per year to obtain accident frequency, i.e. the expected number of flooding events i per reactor year. The number of valve operations relates directly to tne number O of maintenances perfonned per year since the isolation valve musc be closed and reopened once during maintenance. Hence the units on event A should be the expected number of on-line main-3 tenances per year, not system unavailability as computed by SAI. Computation with system unavailability results in an accident frequency about 100 times smaller than with the expected number of maintenance acts per year. When this i point was raised at FRA's July 16 meeting with SAI personnel, SAI answered j with a response described in the attachment. This attachment uses HPCI as an example. 3 J , D-8 J
O TheHPCIsystemunavailabilityis10-2(unitswithoutdimensions)andits maintenance frequency is
.09 acts , 12 months ,
1.08 acts / reactor year month year
- O We claim that the expression for accident flood frequency above the six feet elevation concurrent with loss of the power conversion system is .
expected no. probability that probability that probability probability o of on-line the system is dis- the operator operator that the maintenances assembled given opens the iso- fails to operator per year maintenance lation valve reclose erroneously during mainten- the isolation isolates the ance valve power con- _ version system O E[NA (one year)] P(B/A) P(C/A) _P (D/AnBnC) during flood - ing s Where n denotes logical intersection and P(E/AnBnC) t+ At O E [NA (one year)] = A t
= maintenance frequency, maintenance acts per unit time Where wA At = one year t = time i
Again, the reason for the above functional form is that P(C/A) is given on a O probability per maintenance basis. (In section D.6, we develop a probabilistic expression in which system unavailability is a valid term.) Swain and Guttmann give a best estimate of 5x10-3, as mentioned above, but estimate the 90% upper confidence level of P(C/A) as 2x10-2/ maintenance outage. o Becuase the long maintenance outage of 3.5 days (see attachment) results in several shift changes, it is conceivable that the upper bound estimate may be applicable, resulting in an estimate that would be a factor of 4 higher for the O flooding frequency. D-9 O 1
O D.4 Level of Stress Issue O In addition SAI's estimate of the probability of failing to recover, P(D/An Bn C), is 0.05 which corresponds to the non-stressful situation in fig-ure 17-2, Swain and Guttmann. However, if stressful conditions exist, then
.O the estimate is a factor of 5 higher, i.e. 0.25.
Also, one can see from the event tree in figure 3.4.23, that SAI as-signed a probability of 0.12 to failing to restore the condensate system. O Considering the level of stress involved, a probability of 0.25 or higher could be a more reasonable estimate. D.5 Observations Examining the event tree in figure 3.4.23, one can see that $AI calculated a frequency of 4.8x10-7/ reactor year for flooding to the six foot level (T2 initiator) and 5.8x10-8/ reactor year for a class I core vulnerable accident. However, following our arguments given above, the estimate for the T2 initiator could be as high as 1x10-3/ reactor year for flooding to the six foot level and 1x10'4/ reactor year (or higher) Tor the class I core vulnerable O accident--factors of about 2000 difference. These class I accident figures are for the core vulnerable states to be reached; the core r:elt frequency would l l be somewhat smaller, depending on an analysis we have not done. l We recommend that SAI recompute the internal flooding frequency considering that event A should be given in units of frequency. Furthemore, because the internal flooding sequence is dominated by a chain of human error events, we recommend that an organization expert in human factors carefully analyze and lO compute these accident frequencies associated with internal flooding. This recommendation applies to the entire flooding event-tree sequence involving human action, i.e. !O e erroneously removing the power converison system during flooding a failing to provide coolant makeup with the condensate pumps e failing to provide containment heat removal by opening the MSIV's within 10 hours. lO D-10
.7-23 Figuro 17-2 O
from reference 1 O 1.00 0 HEPs
.90 -
g 1.0 - 1st MIN
< .9 AT 5 MIN
.1 AT 30 MIN
@ .01 AT 2 HR o
O bi BUT IF HIGH STRESS PERSISTS, LEVEL OFF AT.25 AT ABOUT 25 MIN. Q a ce - 2 'o S - z IF HIGHLY STRESSFUL Q CONDITIONS PERSIST 5 - - - - - - - - - - - - - - - - - - - - +
$ . 25 -
M O - AUTOMATIC SYSTEMS WORKING OK
.10 -
.01 II I ' i 15 10 30 60 120 0 TIME (MINUTES AFTER LARGE LOCA) lO Figure 17-2. Estimated Human Performance after a l Large LOCA l l
O l O l D-11 O
D Also, as described in the main body of the report, an important assumption made by SAI is that flooding to the six foot level will not result in automatic D closure of the MSIV's. (SAI does assume however that reactor trip will occur). It is important to verify that the assumption regarding automatic MSIV closure is true--otherwise the power conversion system is lost and the only normally available coolant makeup system is the condensate system. In this case the S accident frequency caused by flooding would increase by an additional factor of about 10 and design changes may be necessary to mitigate the effects of the internal flooding accident sequence (e.g. elevating the ECCS pumps to a higher level). O S S D D D D-12 9
O D.6 Alternate Probability Expression We can develop an equivalent expression for flooding frequency in which 3 system unavailability is a valid term. In this case we must know the conditional probability per unit time that the operator inadvertantly opens the valve during maintenance, defined as A below. C/A O For notation let Q denote system unavailability P probability A frequency, probability of occurrence per unit time g n logical intersection The accidental flood frequency above the six foot level per year concurrent with loss of the power conversion system is O t+ A t f QA P(B/A) P (D/An Bn C) P(E/An Bn C) A C/A dt S t d where At = 1 reactor year S is probability per unit time that flooding The term QA P(B/A) P(D/An Bn C) AC/A will occur above the six foot level. The term QA P(B/A) defines the fraction of time the system is vulnerable to flooding. We can calculate A C/A as
' Probability operator -
inadvertantly opens Maintenance talve outage C/A " maintenance outage - time 9 -
= 5 x 10-3 . maintenance outage 1 day maintenance 3.5 days 24 hours outage D =
6.0 x 10-5 / hr It must be noted however that the approach outlined in this section would still result in a factor of about 100 higher estimate than SAI's estimate for the y flooding frequency. D-13
b O O REFERENCES O-
- 1. A.D. Swain and H.E. Guttmann, Handbook of Human Reliability Analysis -
with Emphasis on Nuclear Power Plant Applications, NUREG/CR-1278,
- p. 13-4, April 1980 0
0 3 D 0 0 D-14 0
.O O -oba 01 - Attachment In %e., f\oedi eventsc4aloo which kion oC %e 4re90eg of inkenal could \eod 4o potenkial ,O vulner ble, o.oeditions, bhe c.ote op ;ngernag gloodi re.lo.M onship of thn 4'rgveg on cer4ain sys+c cuents 40 imhor- y^oin+enance ac.As
'ts ecquiced. Expla'sn 4h'is retoAionship.
cr '""""'"
$ On-be mLwkenanoa, o4
' iE g remekr boud's wq . h)hich ca.n sysiems rehaase, ioco4ed in 4-h-c ra +he,eene.br bdW6'i d er ink Me. ' i 6;ni be. +he, dominan+ m has been e.coloaled since W may-
.d fobuna me.W1.cen46bu%c is used
% okoMa\ ficodig. The.
4o oC vonftfy 4he Ocequency
}g*j i ofia\ ma'tn+enance. ere es \eodt t to +he c
te e. d4 usakee 'tnio On-teac.4cr iIdiq. O Ca\co\abe, %e being subject 4o the, covar%owat telease o4-pobE\Mj d ibe. system b t\M a ConcGon of 4he t. cot 4,ec -6 4ha eeocher Q e9sM.r c4 +i me, i+ 'is O Mainkno.nce h l.haua'iiabui4y)I 6AV.hX 5 I[FracNion o4foh u sysun sc Aslem W 4., up o. h Ona*d6s 5 5FL ms. uun a i k ml14oman Becots o ccur O (Tokle. C.8 SRs DRAY) fFA TMs can be_ dewed - in htms o9 a. demand cbbW+y 'O *in We. fokung con 4 ext 4 hose, events, for c.h m:6n-4enance a\\ reguires opedir$ We sys+em up fee mo.fce cep.'tes in W.Wod regutces vnuch we. kn We. enean ftme, of repairs of 4ha, associad un% a.i.L. odakeance. act-s. *
- Because. .
! sysicms the sign'iOcard mean eWovb 'snebed \n esassemerg O approis m,oh.g 4ime 4'oe %ese, oc+s, is asknn4ed be 3.s' day (i.e. ,) W. We tec>sAco.t speci4co. tion limit 4'or on-be < dinierece .
* '%e. Aqye of such everds o- which is be. ep. Is Enh 4o be kss %an annmh
" usR- gqoo ack Ao b e.vatohed
- n. p sG m c. %e.k ly tcic^ean kweshe. . of an. winkenance
'O D-15
I 1
' h**S am Wie c4coiaMen ad h W ocauditaWig l
'O ' of Wg cys%m, i+ is Sourd %at for M9c.2 coch oe.%5 occur uk+h a Seeqpenoj e4 opo4mately:
- I"#
O E'%***1 Mo 65 Y 3.s % s/ $ q . g \ .09 acts ljj Wequene.y 1 $e;ts/ Mo. System) = mo. -syelem owc 3D L
't G '
- 3 i An 'impt.ch assumpbn in %ts exaica4 ion is %a.t %e Q!
'. frequency of such rnaioc on-rine main 4ecance ss apo>iirndely once. pec y eat, a+6 Weebe bhe "e'isW' asseeloNed wi% EVis on-rine eminknance is Asc.ewj propoe&,som\ +o we. legth e e, acAvo.\ of +be fe s ooge,4- such maior on-fine work effee-Esis.e O si ki oenek.n er % e. used he.ee nece, ce$sticoatly ta be aMs and con %buWn b -Ceequendee. Gl<a.Ghood $G a. te\eose due, % we .h'igher sobsVan4'gVeegue tanyof- such events. 4r vency SpeeW oa\ty, if %ece. is .a. %gber
^ ^*' "" "'* '
'O . sys+""bems*y then % 4. chance, of misopera. Vion doeing 94 pt,-re e t eiocnt g t h e. steen -b s.ervice cootd produce o. w- , h D 'sblo co64NE ' an* Ot4444e, I 4 'tS 'iodged M M pro ' U kj 44 a t el ease,Mof- wakec- 40 a. o+gvan e.lsuaVton s5 dieccMy propoc4Aonal % e he.xqth M fo ' W o *. 4.ite.,
- O gj3pe, .s o enad op o.nd sobjeck %:> %a pos4ckded t
- , fat tote. mocis.
O . l . .
. 9 O
0 . D-Iti , O
- = - - - -= ^ -
t I l l BRAFT-PRElJENARY ! The purpose of this section is to define the best estimate values used to g quantify the fault trees for evaluating the conditional probabill,ty of l muintenance operations leading to the release of large quantities of water - it.to the Elevation 8 compartment. ( ( l
- 1. The probability that the ECCS components in Elevat'on 8 may be lO disassen61ed during plant operations. TMs conditional probability l must be less tun the overall on-line maintenance unavailability taken from Appendix A.4 and shown in Table G.8.
l Table G.8
'O MAINTENANCE' UNAVAILABILITY TOTAL STSTEM CmetTIONAL PectA81LITY .
sTSTDI UmAVAILASILITY OF ST5fD4 CPthED Dua!MS (AP9EPets A.4) PLANT OPCLATIQu (ESTIMATED)
'O l Loop A I a10*3 2 a10** '
Laos 8 2 a10*3 a sle d
) LPCI
% Leg A 4 a10*I O' g -
Pune Leg C PumpLeg8l a a10*3 0 *II Pwe Les O MpC1 10*I 1 a1G*3 RCIC 1.la10*I 1.1a10*3
- O mg ,, 3,3,s -s
* */m" Mar.*"/"lH4::'rd:::." '*** * * .
The conditional probability of the system being opened is based O upon the following considerations: i e only a small fraction of the maintenance operation involve opening of the system to the Elevation 8 . atmosphere; therefore, for most system maintenance .O
, the system is not subjected to the operations fatlure mode(90%) of Interest, i.e. internal flooding
. of the Elevation compartment.
e { D-17 .
- O'
J-O DRAFT PREUMNARY ( e A portion of the maintenance operation is assumed c
- to be involved in disassembling and assembling the O components; therefore, the system is not opened during this time to the Elevation 8 and also does not contribute to the potential for water release. ,
- 2. The conditional probability that the operator opens the isolation valve to the system while it is under ing maintenance is O 'termined from Swain _gnd_ Guttman _
. The value is dete'rniined toK5x10-J/ valve operat _ This error rate is higher than normal plant operatoi error rates since it has been shown that operator error under maintenance conditions may higher than that found performing normal plant function.
!O G.5.3 Operator Failure to Take Appropriate Mitfoating Action In addition to these input quantities, there is another vital aspect of the quantification, the time dependent operator intervention error rate. O Figure G.17 displays the error rate used to characterize the control room operator response to a rising water level in the sump tanks and subsequently in the Elevation 8 compartment. f $( W O # -
- 1. 0-I st Min
.9 at 5 Min
.I at 30 Min lq 3 g .01 at 2 Hr E
W E O u-g; ,, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - : x
,c-3 _ Automatic systems working ok
$ .a l' '
a i lg- g u is l 0 TIME (MINUTES AFTER INITIATOR)
' ' Figure G.17 Time Dependent Operator Error Rate ,
Applicable to the Operator Desnonse N- for Internal Flooding in the Elevation 8
,g Compartment.
D-18}}