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! g NUCLEAR REGULATORY COMMISSION
. 7, p WASHINGTON, D. C. 20555
(....,/ NOV 2 61984
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MEMORANDUM FOR: .Themis P. Sli is, Director Division of Safety Technology, NRR Ormon E. Bassett, Director Division of Accident Evaluation, RES Frank P. Gillespie, Director Division of Risk Analysis, RES FROM: Robert M. Bernero, Director Division of Systems Integration, NRR
SUBJECT:
PROPOSED COMMISSION PAPER ON THE REGULATORY USES OF SOURCE TERM RESEARCH The NUREG-0956 is now scheduled to be published in early 1985, and will sumarize the radiological sour,ce. term research results.
Upon publication of HUREG-0956, we anticipate that licensees may seek changes in their Technical Specifications, or exemptions from criteria in many areas that relate to the new source term, such as filter testing
.and efficiency requirements and containmerit integrity requirements. The potential scope of these requests could be broad since many of our Tech Spec limiting conditions or testing requirements are based upon a radiological source term. While such changes may be appropriate, there will also be a change in emphasis to a more risk based approach to regulation. Specifically, while relaxation in some areas may be appro-priate, our, understanding of accident risks indicates a need to emphasize other areas such as containment integrity. In order to handle this potential influx of requests in a manner that is both technically consistent with the material reported in NUREG-0956, and consistent in implementation, I propose to prepare a Commission Policy Paper to.
accompany the forthcoming NUREG-0956. The Policy Paper would outline
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the manner in which the new radiological source term would be handled in the licensing arena. The Policy Paper would thus explain to the public
- and the industry how the Commission intends to utilize the research results, to scope the areas in which changes would be accommodated, and to identify the process to be used to effect any changes.
The Policy Paper, to be comprehensive, should represent the thinking of a brced constituency within the staff. I request, therefore, your
! designation of kncwledgeable first line managers (preferably a Branch l Chief or Section Leader) to participate in the development and l coordination of the Policy Paper. I suggest that participation include representatives from the Fuel Systems Research Branch (ASTP0), Reactor l Risk Eranch, Reliability and Risk Assessment Branch, and three branches l in my Division (the Accident Evaluation Branch, Containment Systems l Branch and the Reactor Systems Branch). I suggest all activities, l inclucing review of early drafts, :hould be coordinated through the line 8507130437 850426 .
BNA1 199 PDR < I L
- - . _ _ . _ _ .. l Multiple Addressees organization to help ensure a staff consensus before final management review.
The draft of the paper should be scheduled for senior management review by March 1, 1985.
I have asked L. G. Hulman to coordinate this effort.
Please inform Mr. Hulman of your nominations, or me if there are conflicts. Similarly, by copy of this memo, the AD for Reactor Systems should identify the Containment Systems Branch and Reactor Systems Branch representatives. .
A statement of objectives and considerations for use in drafting the Policy Paper are attached.
Robert M. Bernero, Director Division of Systems Integration ,
Attachment:
As Stated
.cc: H. R. Denton R. B. Minogue R. C. DeYoung E. Case D. Ross T. Rehm E. L. Jordan R. Vollmer D. Eisenhut H. Thompson G. Arlotto K. Goller D. Muller L. Rubenstein F. Rowsome .
M. Silberberg A. Thadani W. Minners B. Sheron W. Butler D. Mathews J. Malaro W. Pasedag L. G. Hulman
CHARTER FOR MANAGEMENT GROUP TO DEVELOP A COMMISSION POLICY PAPER ON INTEGRATION OF THE RADIOLOGICAL SOURCE TERN RESEARCH RESULTS '
INTO THE REGULATORY PROCESS
- 1. Become sufficiently familiar with the ASTP0 sponsored methodology, APS review, WASH-1400 source terms TID 14844 use and related risk considerations, to judge policy and practice implications;
- 2. Identify areas of existing policy and practice influenced by source tenn research using NUREG-0771 (FOR COMMENT) and the attached sunr.ary memo as a starting point for identification (note that both NUREG-0771 and the attached provide such a profile through 1981, but require updating considerations in such areas as severe accident rulemaking, safety goals, safety issue prioritization and resolution and emergency planning);
- 3. Identify future policy and practice areas potentially influenced by the evolving source tenn research; e.g., proposed safety goals, siting, environmental statements, severe accident policy, and DL Task Force on Tech Specs.;
- 4. For each area of regulation that is influenced by the research that may or obviously will have a significant effect on regulatory practice now in place;
- a. describe the current regulatory requirement or practice,
- b. make a tentative evaluation of how it might change based upon new source term information, c.- make a preliminary estimate of costs and savings achieved by b; above, and
-d. make a recommendation on whether to pursue regulatory change in the area and how it should be done.
- 6. Provide a plan or policy statement for evaluating
. applicant / licensee requests for deviations from present practice.
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BROOKHAVEN NATIONAL LABORATORY MEMORANDUM DATE: September 27, 1984 To: W. . r t CJ ,
FROM: K R. er ins and S. Y. eh-
SUBJECT:
Preliminary Review of the Containment Response Analyses in the Shoreham PRA INTRODUCTION We have conducted a preliminary review of the containment response analy-ses contained in the Shoreham Probabilistic Risk Assessment.1 A parallel ef-fort sponsored by RRAB/ DST /NRC is under way at BNL to review the event tree developnent and quantification. This " front end" evaluation is a much more extensive review than the present review and it has provided many valuable in-sights. We have concentrated our review on comparisons to Limerick since BNL has gained extensive experience in the previous review,2,3 and the plants are very similar. Our review has thus concentrated on areas where there are analytical differences between the two PRAs or containment design differ-ences. The degraded core frequencies for the four accident classes are shown in Fig. 1. The definitions used in the Shoreham classification scheme are in-cluded as Attachment 1. The dominant differences between these two estimated frequencies is in the Class IV ATWS with Shoreham being two orders of magni-tude higher than Limerick. Class II loss of containment heat removal is also estimated to be higher in Shoreham than Limerick. In order to keep this com-parison in perspective, a comparison of the results for all the available BWR PRAs is shown in Fig. 2. Note that the Limerick PRA gives substantially lower core melt frequencies than any of the other PRAs. However, methodological differences make direct comparison between the various PRAs difficult. The Limerick PRA used the basic approach and techniques of the Reactor Safety Study (WASH-1400) but accounted for plant specific design differences between Limerick (BWR4 with a Mark-II containment) and the WASH-1400 plant (BWR4 with a Mark-I containment). The Shoreham PRA methodology is compared to WASH-1400 in Table 1.
The high frequency of ATWS events in Shoreham is of particular concern because of the potential for severe releases. Much of the difference in ATWS frequency can be attributed to the lack of an automatic poison injection sys-tem and to differences in the ADS inhibit logic. In other respects the scram systems used in Shoreham and Limerick are quite similar.
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- W. T. Pratt September 27, 1984 Having noted that the ATWS core melt frequency in Shoreham is relatively high, it is interesting to note that the calculated radiological impact is only moderate as indicated in Table 2. This is basically because of the large decontamination factors (DFs) calculated for Shoreham. Because of the differ-ent classification schemes used in the PRAs, it is difficult to make direct comparisons of the DFs. However, in order to get some perspective on the high DFs claimed for Shoreham, the release fractions for a typical ATWS sequence are compared in Table 3 for the three plants. Both WASH-1400 and the Limerick PRA calculated severe releases for these rapid sequences, but Shoreham calcu-lates releases two orders of magnitude lower. We noted above that the Lim-erick PRA used WASH-1400 methods so that one would expect the source terms predicted in the two studies to be similar for compatible failure modes. How-ever, from an inspection of Table 1 (Item F), it is clear that the Shoreham PRA used more recent methods to determine the radionuclide source terms and most of the reduction is apparently due to higher pool DFs. It will be im-portant to verify that these reduction factors can be achieved under all se-quence conditions and failure modes.
DISCUSSION .
Of the five Shoreham plant accident classes, the final set of risk con-tributing accident sequences are chosen based on the ranking of importance of the product of the end state probability and source reduction factors. Four-teen risk contributing release categories and two non-risk contributing cate-gories are defined. Three of these sixteen categories are Class IV accident sequences. They are SNP-10 (CgRg T3 -y), SNP-11 (CgRg T i -y), and SNP-12 (CgRgT g T -y") as described in Table 4 Note that SNP-12 consists of both y;-y';
andCy'Rgscenarios where the y" scenario assumes the wetwell failure be-i low the waterline at the basemat. The DFs were calculated for the fourteen categories as shown in Table 5. Among the fourteen categories, three are Class IV accident sequences. However, the y" sequence was not included in this table. The pool scrubbing DFs for the various accident sequences are summarized in Table 6 with the implication that the DF for y" sequence are at least as much as the values in this table. The high suppression pool DFs of the Shoreham plant are based on the assumption that the pool is intact, and all fission products go through the pool [with the exception of Class IV (CgRgT i -y) sequence in which 10% pool bypass is assumed] before entering the containment. In order to evaluate its high DF claims, the containment struc-tural design of Shoreham and Limerick were examined and compared. The follow-ing preliminary assessment can be made.
- 1. The diaphragm floor at elevation 62'8" of the Shoreham plant was not an-chored to the containment wall as in the Limerick design. The Shoreham containment wall displ acement will expand outwardly under pressure as shown in Fig. 3. Based on this free standing diaphragm floor design, the PRA suggests that the most likely leakage paths will occur at the junc-tion of the diaphragm floor and less likely at the basemat. Under this
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'9- J-W. T. Pratt. September 27, 1984 failure' condition, the suppression pool integrity and high scrubbing ef-ficiency will probably tHe maintained. However, when gross containment failure occurs, as will be the case in many Class IV accidents, the base-Lmat-containment wall juncture was judged by Stone and Webster as the most probable location to fail (Appendix M of Ref.1). Under such failure conditions -(generally defined as the y" scenario), the suppression pool water will'be blown out into the surrounding chambers.
- 2. The suppression pool of both plants is surrounded by chambers; while the Limerick surrounding chamber is partitioned, the Shoreham surrounding chamber is a continuous annular-like space. Both surrounding chambers have drains. Limerick's PRA assumes drainage of. the suppression pool in y" sequence. Likewise, it 'is reasonable to assume that the Shoreham sup-pression pool will- also be drained under such failure conditions (y" se-quence). If such'is the case, ~the DF of Class IV y" sequence should be evaluated explicity.- At present, the Shoreham PRA does not include the y" scenario in any of the sixteen release calculations. Instead they are
" binned" with the y' sequences where the pool remains intact.
- 3. ' After the bottom head' failure, the Shoreham PRA predicts that 90% of the core debris will flow to'the suppression pool via the four downcomers un- -
derneath the vessel -in the CRD room. The remaining 10% of the core de-bris will attack the concrete floor of the CRD room. Because of the lim--
ited amount of molten corium, the core-concrete interaction does not gen-erate a substantial amount of gases to threaten the containment
' integrity.
The estimate of 90% of the core debris flowing into the pool is probably a very good estimate if the molten core debris can be treated as non-vis-cous incompressible fluid (as modeled in' Appendix L of Ref.1), since the remaining molten core on the concrete floor cannot be more than 1/2" deep before it spills over the downcomer's neck and flows into the pool. How-ever, there is a wide range.of possible debris conditions at the time of vessel failure." Generally the high temperature molten debris (~4300F) is taken to be the limiting case. For Shoreham, however, the low temper-ature solid ' debris (~2700F) ' may be the worst case since very little ~de-bris would flow through the downcomers. Thus the effect of more than 10%
of molten core remaining on the concrete floor should be addressed.-
The revised geometry (see Fig. 4) of the downcomer vent pipes is intended to maximize corium flow into the pool but this also increases the poten-tial .for steam spikes and oxidation release.
INFORMATION NEEDS In view of the previous discussions, we.would like to request the follow-ing information:
- ~ -
9 8 W. T. Pratt September 27, 1984
- 1. The revised reactor pedestal geometry is not described adequately in the schematic (Fig. 4). Verify that the vent pipes and manways remain un-blocked in the revised pedestal geometry.
- 2. Provide the estimate of the fraction of molten corium which is expected to spread out of the pedestal area through the open manways and vent pi pes.
- 3. Verify that the downcomer vent pipes only protrude 1/2" above the dia-phragm floor of the drywell as indicated in Fig. 4
- 4. Section 3.6 of the PRA takes credit for containment leakage which will prevent gross containment failure for all pressurization rates except the very rapid pressurization associated with large breaks. However, the structural analysis by Stone and Webster (Appendix M) did not identify any significant source of leakage. The expected leakage source and the leakage rate as a function of pressure should be provided.
- 5. The basis for the partitioning between release category 10 and 11 (no pool bypass vs. partial pool bypass) should be provided. The phenomeno-logical basis for the estimate of only 10% bypass should be provided.
Preliminary results from IDCORE indicate that for some BWR sequences the vessel will fail with only 20% of the core molten. Presumably 80% of the melt release would bypass the SRV's and be released into the drywell.
- 6. The basis for the binning into release categories is poorly described and the transfer from Tables H.4-8 etc. (Attachment 2) into the 16 release categories is inscrutable. A table listing the specific sequences which are binned into each category should be provided.
7 The lack of R5 sequences in the release categories makes it apparent that these releases have been binned " downward" into the lesser release cate-gory Rg. The basis for this " downward" binning and any other sequences that are moved to less severe categories should be provided.
- 8. Table H.4-25 appears to be incomplete in that it does not include se-quences 06 and D8. The completed table should be provided.
- 9. The source escape fractions used for end state screening (Table 3.6-10) appears to be quite arbitrary yet it greatly influences the importance ranking. In particular: the use of Z as the surrogate for melt release ignores the fact that there are noble gases in the melt release which will not be scrubbed at all; the use of a large scrubbing factor (500) for Cg transients is inappropriate since most of the melt release will be released directly to a failed containment; the reduction factor of 0.01 for y" failures is indefensible since the event tree precludes everything but large ruptures where the pool will be blown out into the reactor l
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- W. T. Pratt September 27, 1984 building at high pressures. Taken in conjunction with the scrubbing fac-tor of .002 the reduction factor of 0.01 implies double scrubbing with a decontamination factor of 50000 for an event in which the final level of water in the suppression pool is highly uncertain.
Table 3.6-10 should be replaced by a table with defensible reduction fac-tors. As a minimum the table should include a separate category for Cg transients, which recognizes the defined sequence of events (containment failure before core melt). In addition, a detailed justification for each reduction factor should be provided along with the numerical results of the ranking process.
- 10. Sheet 1 of Figure H.4.2 has been reduced so that it is illegible. A full-size legible copy should be provided.
RECOMMENDATIONS l
In addition to the above information, we feel that there are several l
areas which are important enough to warrant independent verification. The ba-sis for our concern and the proposed resolution for each item is outlined below:
- 1. Core debris disposition: The partitioning of 90% of the core debris into the wetwell is highly speculative and assumes that debris will be nearly inviscid. In fact the molten core may be very viscous and may be solid-ified by quenching in the lower head of the vessel or on the drywell floor.
We propose to run a Class I accident sequence (e.g. TQUV) with 50% of the debris retained on the drywell floor in order to examine the potential for early release of fission products for this class of events.
- 2. The Shoreham PRA presents no quantitative analysis to preclude failure of the wetwell below the waterline. In fact, Appendix M indicates that the most likely failure location is at the bottom of the wetwell. A failure in this region would force the pool into the annular region of the reac-tor building surrounding the primary containment. If the reactor build-ing does not fail and the drains are not on, the pool may still tend to mitigate releases from the containment as they are bubbled through the failure location into the reactor building.
We propose to use SPARC to address the DFs for the y" configuration as-suming the pool is retained in the annular area surrounding the contain-ment. We will also assess the significance of the assumption of no pool DF for this sequence (as assumed in the Limerick PRA). The partitioning between y" and non-y" scenarios will be based on the applicant's response to questions 1 and 2 and the structural analysis of Appendix M. The l
4 &
W. T. Pratt September 27, 1984 partitioning between scrubbing and no scrubbing for the y" scenario will be based upon our assessment of the possible pool configurations after a large rupture at the basemat at high containment pressure.
- 3. The release fractions for Shoreham are several orders of magnitude lower than both WASH-1400 and Limerick. Most of this difference can be attri-buted to high pool DFs based on limited experimental data as cited in Ap-pendix N.
We propose to use SPARC to assess the potential for lower decontamination factors for a range of conditions. The calculations will emphasize va-porization release to a saturated pool since our previous experience in-dicates the potential, for lower DFs under these conditions.
4 The issue of the energetics associated with steam explosions remains un-resolved, but the issue is being addressed by the NRC as part of the Con-tainment Loads Working Group (CLWG) effort. We propose to review the preliminary results of the CLWG effort to ensure that the NRC position is consistent with the low probability of containment failure (4x10-4) or the low probability of an oxidation release given containment failure (6x10-3) that is cited in Appendix L of the Shoreham PRA.
REFERENCES ,
- 1. Science Applications, Inc., Probabilistic Risk Assessment, Shoreham Nuclear Power Station, SAI-372-83-PA-01, June 1983.
- 2. H. Ludewig, J.W. Yang, and W.T. Pratt, " Containment Failure Mode and Fis-sion Product Release Analysis for the Limerick Generating Station: Base Case Assessment," BNL Informal Report, BNL-NUREG-33835, April 1984
- 3. I.A. Papazoglou, et al., "A Review of the Limerick Generating Station Probabilistic Risk Assessment," Brookhaven National Laboratory, NUREG/
CR-3028, February 1983.
- 4. " Reactor Safety Study--An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," WASH-1400, NUREG-75/014
- 5. K.R. Perkins, G. A. Greene, and W.T. Pratt, " Appendix D-Standard Problem 4 (BWR Mark 1)," Appendix D of the Containment loads Working Group Standard Problem Results, to be published. .
KRP:sm/tr cc: R. A. Bari G. A. Greene D. 11 berg H. Ludewig hJ. W. Yang R. Youngblood W. S. Yu
y Table 1 Major Changes in the Shoreham PRA Compared to the WASH-1400 Methodology
- a. New sequence initiators are defined and accident sequence models developed, including time phase event trees. .
- b. The definition of generic accident release categories in WASH-1400 required lumping accident sequences with major differences in potential consequences and containment interactions into the same category for consequence evaluation. For the Shoreham evaluation, realistic and refined release categories are defined so that each unique sequence type could be evaluated separately assuring greater detail in defining the spectrum of radionuclide releases,
- c. Smoothing of probabilities among release categories was used in WASH-1400 to account for possible miscategori-zation of sequences and other uncertainties. This artifice is eliminated in the Shoreham evaluation because of the better definition of accident sequence release categories for consequence evaluation, e
- d. Accident sequences are totally reevaluated using the latest BWR thermal hydraulic calculations for trans-lents. LOCAs and ATWS.
- e. Component failure rate data and common made fail ures are reevaluated based upon the latest data from operating nuclear plants.
- f. The radionuclide source term, release mechanisms, and removal mechanisms have been completely reevaluated to incorporate the latest experimental data and analytical methods in the characterization of source terms.
- g. The conservative estimates of the probability of the steam explosion leading directly to containment failure was ' reassessed. The steam explosion phenomenon leading directly to a containment failure and substantial oxi-dation release is realistically evaluated considering the specific Shoreham design. The probability of this event is reduced.
- h. The conservative assumption that all potential . core damage, sequences lead to a major release sas reassessed. Detailed containment event trees were developed to appropriately characterize the accident sequences which could lead ,to a radionuclide release.
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Table 2
SUMMARY
OF sigil!FICAtlT RADI0flVCLIDE ItiVEtiTORY
. RELEASE FRACTI0 tis BY RELEASE CATEGORY actlesse (seenssase ,,,,,,,,, ,,,,4,,,
est raag nestil euteret assinesssas y
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. (%) ?>, P*sture of !ste9tt31 facett it a relative Peature of raetele19481 delet te se eseeted lettvidwel.
e levere *atat tPat t*e aelle Setet seg tusttsattel ***ttleat Of t*, volettf et dad atttete*t are F91*elee.
. **terate eeat tast t*e aosle ttles aae t-ell 'rettttas of tae valettlet aae attinteel are relettet.
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i TABLE 3 V .
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Comparison of Release Fractions for '
ATWS Sequences s s
.g Iodine or Cesium Tellurium Shoreham <
Release WASH-1400 Limerick Shoreham WASH-1400 Limerick Shoreham '
Category -E .s 11 .11 .07 .732 1.6x10^ .3 .1 .762 2.5x10-2 I Based on comparable BWR release assuming elemental iodine. ' -
2 Depending on failure mode. ]'
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Table 4 . -L DESCRIPTION OF Tile RELEASE CATEGORIES IDENTIFIED FOR Tile Sii0REllAM PRA (Sheet 1 of 4) ,, _
T attfait goethAef ACCIO(hf CIf CAltLour GEN (RAL M5CRIPTI0ll ,- SIquENCE CoulAIBui!04 P00G(15104 PAtu '
BASIS f04 Is.PtAnt AnatTSIS O(110 4104 I
i ihr$.3 This release category is representative of Class 1 accident sequences less of Off. site Power taltlater. failure Cg4e 7, 4 l levolulag a ers steet event leadlag to core meltdo a where the coa. to recover Oi, Isles I er 11 electric pe=er, talancat falls to tselste er falls by everpressure early la the accl. failure of halt pressure lajectlen systems. Early f ailure la 1 dent se+.cate le4Jiat to a leakage type release from the drywell. failure of A05, failure to esetate la the the Or3wil drpell. , 1 i
t A 9 Stips.2 This release category is representative of Class I accident seguences RS flece utlch partially dralas supprestlea involulag a transicat event where the contalascot falls to isolate er pool. failure of high and 10. pressure la. Cg 8,Tg.3 falls by everpsessure early, leading to a leakage type release from jectlea systems. failure te Iselate la the Early faltere la the wetucil. metuell. the wetuell i.
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t Stirs.) This release category levelves a core seltdown for a Class I acci- toss of condenser vacuum. failure of high Cg h,fg-7 dent sequence la which the contalement f alls la the leag tern lead
- pressure lajcClion systems, failure of A01 Ing to a leanage path from the drywell. The long slae to costalement the costalament is f atact during the sig. Late failure fallure is rapected to reduce the alvinerne finsten products la con- alficant fisiten product release perleds. la the Drywell 1 talaavat substaattally prior to release. ,.
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$hP1 4 This release category levolves e care meltdown for a Class l SCct. Less of cendraser vacuum, failure of high dent segumace la which the centstament f alls le the long tern lead. prrssure lajection systems failure of 801 gggI 4 4 . y.
lag to a leasese path frees the estuell. the long time to contalament the coatatament is tatact durlag the sig. Ceiggg=P failure is espected to reduce the altborne fisslea products la coa. alficant finsten product release perleds, talmeent substamt telly prior to release, tale failure la the sletuell 4
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DESCRIPTION Of Tile RELEASE CATEGORIES IDENTIFIED FOR THE SHOREHAM PRA (Sheet 2 of 4)
SfliAlt 80MthAh! ACClethf CfF CAltcuar CthfRE M 5CRtrilen stQuthCC Confalgulleit P90ca(15I03 Patu SA515 FOR IN-PLAh! AAALV111 015tC3A104
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She5 5 thls case,or,la.ei.es Cl cataia~at f seis e t. mi mn of uc*ca
- at,f. se,en.es u se ca,ahiilt,
. hub ihe am of c.aame uc . fan res .f ad. Ce,r 2 I .,
eat .ai at:CSC. aad PC14 fati re i u iant lasec. +
f.n e4 h,c -e .eitde a. ihe cota;.u at finne $s moaa tia f.il-lag catalaneet fanwet she f a n.e la le accur la the drywell and the fissim products althorae la suppresslam peel is effettlee la retalalag the Dry = ell !
costala~ at would eet be SIgalficantly redwced, the fisslea products released from the c.re reis se.. m ei. ,
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- I Shr5-6 This categu, lo i.es Cias l
c ista cat fans e i. i.s itisucleat
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. huh the im .i c.amme m t fan-es .I ed.
at:CSC. and eC14 feli.re .f ca.iant lajec. C,s,r8 . y ,
f.ne.a h, cue .eitde a. rh. c tain-at fan-e is momed to occur la the wctuell aad the fisstaa products althorse la ti.e f.ne. leg catalantal fan,et the sopprelaten pool is effectfee le relatalag
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coatalement would met he significantly reduced. the fisslea products released free the g' U '
core regnea la. vessel, t% O"rywen"
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ShP5-F Ibis release category is representative of a Class !!! accident LOCA. f4Here of esper suppres:I**. I tar ear {18'erpressure failure of centataarat. C3 8,Tg .7 sequence la which the coatalament falls early la the accident segueace due te leade.luate pressure supprestlea capahlilly. The Ierlr faltere fission products seleased free the core regle* are discharged la the cryme n directly to she drywell ateesphere and are met sigallicantly 6 attenuated prior to lea 6 age free the dry = ell. This category ,
lastedes tww 10CA and hPV lallere accideat sealmences, which challenge contalmeent tategrity early la the seguence.
1hr5-8 Ibis release catepry In. elves a CIast lll eccident sequence la im . fanure of high and law pres. C/,T, .1 unich the contada at salts la the long tena leadlag to a leasage ** I'l ' '** ' #5 "S * '"* * *"8 8 8""" E 8 8 path frees the day = ell. Ihe lang time to centalmewal f allere ls '*'# "'"I I ' ##458""'8Id8 !'888 I 8I' I8 0IW8 espected to redwe the airterne redleauclide material leventory '"I'd I"** '"*
' '"'I* 8" *** 8'F"'33 la contata== at prior to its leabage to the eastreament.
l .
. I
.i
Table'i - -
DESCRIPTION OF Tile RELEASE CATEGORIES IDENTIFIED FOR Tile Sil0REllAM PRA (Sheet 3 of 4) eftfA5[ 00MthANI ACC10thi (If CAlt(Mtf .
GihteAL 0(5CeltTIost stQuthCC Conlettuttosi ptocef5510s PAlu BA515 f04 lu-plant ANALY515 CE58f.1 AIDS 1Nr5 g This release category favolves a Class lit accident sequence la stedlue LOCA. failure of high and low pressure C3 e,T, . 7
- which the containment #411s la the long tern Isadlag to a leakage lajecties systems. the contalament is latact path from the =etwell. the lang time to costalament failure is d*' lag the radioavCllde release perled from C)e 14 e .1=
enpscted te reJace the altborne radicauclide material laventary I"* I"*I* tale failure la centstamcat prior to its leabase to the environment. la the wet. ell 1NPS-10 This release category is representative of transleet events It5lf Closure ATV5, fatture of StC. failure lavotulag Cists I. Class 11. and Class IV accident sequences where of all lajecties systees falleulag contata. C,e,T, 4 the suppressten pool la partially bypassed and the contalement meat f ailure.,105,ef the fisslon products lategrity is test early in the accident sequence. released from the fuel'In-wessel'Is (reas.
g,,,,y,gy,,,
ported directly to the drywell bspasstag la ene Orywell the suppression pool.
l l
Milt Closure A!VS, failure of SLC. f alture
~
1hr5-Il This release category levelves e Class if accideat seguence la C,e,Tg.1 which the coatalament falls by a failure to scree and remove of all lajectica systems felle lag contala-decay heat, followed by core seltdova. The containment failure meet f ailure. the fission products released free the fuel la-vessel ace totally trans. far fd M is assumed to occur la the drywell and thc_fjssjon goducts are la the Orywell sol pof ted te, the contalament through the sup.
leas.3l ent.
E "lfhanti
~ g ttenuated prior to its leakage~ ~to~~**"
liie*ei;] F P'esslam pool.
This release category levelves e Class it accident seq." ace la M51V Closure ATW5. fatture of 5tC. failure C,e,f, . P 1hP5-12 er all lajectlen systems following contale. 1*
which the contale==ent falls by a failure to scree anJ remove mest failure the fleston products released Cet448 decay heat, followed by core meltdown. The contalement failure free the fuel la-velge] art letally trans. g,g g g
is assumed te occur la the wet. ell and the fisslea products are breed re the contalement throuu*Jhe sup-si.jnllicantly attes.uated prior to its leakage to the envirasunent. pressioisioet.-- "- ~~ "" ' *' "' I"' N
~
Table 4 ,
~
DESCRIPTION OF THE RELEASE CATEGORIES IDENTIFIED FOR THE SHOREHAM PRA (Sheet 4 of 4) 80slinAnf ACCittaf Ctf AftfA5E M SA1. W alpfl0E $(quthCE Couleleuttes . peerJI(55 ten pAtit ;
cal (Gunt ' '
8A515 70s In-PLAh! ANAlfill 3(SICmaf04 , 3 a
laterfu lag LOCA. the suppresslea peal is CBT
$mP5-13 This release categer is representative of Class V accideng partially effective la mitigating releases. 54I A sequences e.htch lave we core meltdown felleulag a 10CA out.
slJe centstament. The 5AVs are actuated la order to altigate the release of fisslea preJuCts to the entiremment by providlag an alternative path late the centalament (l.a. suppression "
peel) durlag the la. vessel release perled.
1 1
' this release category is representeil of Class V ucidest laterfeclag t0CA. fell.re of Savs. Cets She5.It 3sI -
sequence which lavelve core meltdema felleulag a LOCA out.
- slJe costalement. The 1AVs are assumed met to be opened. .
' aaJ the fistlen products seleased from the fuel totally }
bypass the contalement. 4 I
i f
i i
L
) ShP5 15 This release category is representative of the levolmeted core tess of condenser wecame, failure of high ggg .g ,
melteums accident sequences la which the Centalassat reaalms fatact pressure lajeCilen systems. failure of 405, 354 .t and the release of radleauclides to the environment usuld be very centalement lategrity is salatalmed.
l f small sad dcteralmed by testage to the reacter buildlag.
I i !
i '
less of condenser vacuum failure of high C8T I54 *I i Shr1.I6 this release category is representatlee of the teenlasted core
- aaltdown accident sequences la which the centalament esmales latatt pressere lajectlen systems. faltere of A05 e saJ the release of radleauclides to the enviramment heeld be very coatelament lategrity is malatalmed. '
- sadll. and detesela.4 by testaye to the reacter buildlag, further. i
{
sere. the teleases muald be filtered by the standby filter systees.
- i. , 9 i
i I
9
-- , - n - -
~
s Table f SHOREHAM
SUMMARY
OF DEGRADED {0RE ACCIDENT ANALYSIS CASES s
SECT.T'ict MA!MNT N!MAY 5 p tS5!CN Pcot DESICriATCA FAtutt 515ftM CECCa w tMATICM CC0ts Lit 3 (t)
CESIGNATCA(a) CtPC$tTICN(D) FACTCAS Sat VEMTS CtTggg.8 Cl W T M M M.C CATg4g.4 C!-W T 120(d) 20 G3 M.C Ca7g447 QP-W T 600 100 M.C J
CA7g 4 4 7' CP-W T 600 100 M.C CP W T 6C0 100 M.C Ct241 7 CIT 2 4 g .7' 0' M . T 600 100 M.C CRT3 4 g .7 07-W L . M M.C C47344 7 W-W L . M M.C C3 n T. 7' OP M t - m M.C C44g 2 T A.7 WW T(e) 600 m M.C .
CRT87 44g OP W T 600 100 M.C C2T44g.7 OP M ~T 600 100 M.C C3T.A BP L(f) 200g 100 M.C,4 54g CIIS4l*I (a) CI - Contatnnent Isolation fatture OP = Containment Overpressure fatture CW Containment tresca is in the Crywell (7)
W . Centainrent Breace is in the Wetsell (7')
(t) T . Transtant event with primary system Intact and an effective retentten of 801.
L = LOCA event .ith the effective primary systes retentsen of 23 and IC% for the vasers and aerssels respectively.
(c) M . MA2CM C . CCAAAL a . Canteest vocule of 'A*.A7 1 -
(4) Pool scracatet effectiveaess is recucee sue to the recute1 s w eritace bettet.
(e) Ten percent af tre fission proeucts releases from tPe c:re resten within tre srirary siste1 ts; asses tme pool.
(f) Fif ty pe cent of te,e fission se:cacts retesses frcri sne fwel ts directed into tre canteire=nt ny aseaing all tre SAf 4tscnarse lines curtag care neat up and melteoma. . j l
. . .. -..-.- - _ a . -- -
Table 6.
SHOREHAM POOL SCRUBBING DECONTAMINATION FACTORS PATHWAY EVENT SRV 00WNCOMERS Class 1, 2, 4 3000 100 Class 3 NA 1000*, 100 l
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- CLASS I oOLA JT CLASS 11 MOVA CLASS Ill CLASSIV 1"A"u',. "' %;^M CON TAIN ME NT Figure [ Sumary of the accident sequence frequencies leading to degraded core conditions sumed over all accident sequences within a class.
t
NOTE: In 'the other available BWR PRA's there is no clear distinction between core vulnerable and core melt end states.
3 8 -
6 - sis ROCK 5 - BROWNi POIE FERRY 4 -
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CCRE APP. P) APP. P) APP. P) CORE VENERA8LE YE. BLE CggC yyL3 s gtg AND CCRg CCRE Ana yyg3ggggL,'
CORE PILT vuL* TERA 8LE YULN W ELE CCRE MELT Ano CCR ELT CORE PR T CCR .ELT PUBLISHED BWR PRA RESULTS
. Figure 2 Comcarisen of Frecuency of Core Vulneracle/ Core Melt from Publisned SWR PRAs. i
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b Figure i Reactor Pedestal Schematic Showing Reactor Downcemer Geometry
~
ATTACHMENT 1 Definitions Used in the Shoreham PRA Accident Sequence Classification P:
4 y . . - - -
Table 3. 3-1 -
0
~
ACCIDENT SEQUENCE CLASSIFICATIONS USED IN Tile SYSTEMS EVALUATION .>
At titet h!
}
St(tifg f t (5lG:aAingt PHYSICAL BA515 f0ft Ct A551flCAll0ft SV5ffet Livtt CONIRISullhG (VLNT SEQUlllCE 8[ pef 5thTATIvt 5(QuthCE (0A CLA55 '.
Class I (Cg ) helatively raplJ core selts containmcat Transients involvlag less of leventory matemps transient with less of latact at core melt and at faltlally owdive and small LOCA events f avolving 58V actu- high and low pressure t 164 pressuses lavolves a release path- atton with less of Inventory makeup; transients coolant makeup way fece the vessel to the suppression favolving less of scram function aoJ inability to prowlde sufficient coolant makeup pool 4
Class 11 (C2 ) Aclatively slow core melt Jue to Iransients or LOCAs involving less cf centasa- Transient with less of I
lower decay heat power; containment is etat heat rconval; leadvertent SRV opening acci- restJual heat removal f allcJ prior to core melig favolves a dents with leadequate heat removal capability eclease pathway frue the vessel to the ,
j suppsession pool i 1
i c.a 1 e t
$ Class ill (C3) pelatively replJ core melt 6 contain-cat ineact at cure melt, but at targe LOCAs with lasufficient coolant nabrup; small sad medium LOCAs with failure of 41.e 1Avs large IOCA with loss of low pressure ICC5 f
saltlally high latasaal peessurai to actuate and lung-team less et inventory f avulves a sel ase feem the vessel sa6eupg SPW falleras with lasuf ficient coolant to it.e Jeywell sabeup -
i-f Class it (C4 ) nelalleely esplJ twee selts cuatals. Iransients involulag less of scram functlen aaJ Translaat alth failure i' ent f alls prior te cure cell Joe to less of contalamest heat teamval er all reattiv- of 875 and failure of everpessa.ees; levelves a release palm. Ity contralg translents with loss of scrae func. St C5 way fee.n ti.e vessel to the suppression lla.s followed 1.y actuateJ Jepressurlastion .I 1
i pool Class V (Cg ) 8.latively replJ care melts contain- LOCAs outstJe containment with lasofficleet 10CA la main stese
= cat f ailed from laltlatloa of asti. coolant makeup to cares laterfacing system LOCAs llaes with failure of Jcat .s e to equis. cat fallveeg involves with lasof ficient coolant makeup itSit closure and loss a selease pathway from the vessel which of (CC5 i Lepenses the chatalan.2nt 6
e l
l i
i
Table 3.6-9
SUMMARY
OF CET CATEGORIZATION DESIGNATOR DESCRIPTION ATTRIBUTE GENERIC ACCIDENT SEQUENCE CLASS C
1 Class I Pool scrubbing prior to transport in contain-ment C
2 Class II Pool scrubbing prior to transport in contain-ment C
3 Class III Pool scrubbing is by-passed prior to transport in contain-ment 4
C4 Clas.s IV Pool scrubbing prior to transport in contain-ment C
S Class V Pool scrubbing and containment are bypassed RECEASE TYPES R
1 Gap Release Recovered accident after initial core overheat R2'N3 Gap and Melt plus Recovered acciden: after oxidation release significant core melting R4,Rg~ Gap, melt, and Un' recovered meltdcwn vaporization accident -
release plus oxidation release I
. 3 417.
_ ~. - _ . - _ _ . . _ _ - _ - - -
a .
- w Table 3.6-9
SUMMARY
OF CET CATEGORIZATION DESIGNATOR DESCRIPTION ATTRIBUTE GENERIC ACCIDENT SEQUENCE CLASS Cy Class I Pool scrubbing prior to transport in contain-ment C
2 Class II Pool scrubbing prior to transport in contain-ment C
3 Class III Pool scrubbing is by-passed prior- to transport in contain-ment C4 Class IV Pool scrubbing prior to transport in contain-ment C
5 Class V Pool scrubbing and containment are bypassed RELEASE TYPES R
t Gap Release Recovered accident after initial core overheat R ,R Gap and Melt plus Recovered acciden 9 3
- after oxidation release significant core melting R ,R Gao, melt, and Unrecovered meltdown 4 S vaoorization accident release plus oxidation release 3iU
. . ~ . - , . , . . , - _ . . . - , . . . . _ , . . . _ . . , _ , _ . - - . _ - _ _ . . . , , _ , . - - -. - - - .
4 l
Table 3.6-9 (Continued)
SUMMARY
OF CET CATEGORIZATION DESIGNATOR DESCRIPTION ATTRIBUTE CUNIAINMENs FAILURE TIME T Time phase T y i Containment is failed before core degradation T Time phase T 2 2 Containment fails during core meltdown in vessel T
3 Time phase T 3 Containment fails during core-concrete-interaction T4 Time phase T4 Containment fails in the long term CONTAINMENT FAILURE MODE BP, CI Containment Sypass Containment isolation or Isolation Failure fails or containment is bypassed 6 Leakage in the Dry- Leakage sufficient te wel'1 preclude overpressure failure B '
Leakage in the Wet- Leakage sufficient to well preclude overpressure failure OP Overpressurization . Containment over-Failure pressure failure, small or large break 7 Gross Containment Drywell Failure Location 7' Gross Containment Wetwell' airspace Failure location 7" Gross Containment Wetwell'below the Failure Location waterline 3 418
Table 3.6-1 TYPES OF POTENTIAL RELEASE FROM FUEL NR No release R- y Core heatup (gap)
R 2 . Core heatup and melt release (gap and melt)
R 3
Core heatup and melt release with ' potential for oxidation release (gap,. melt and oxidation)
R 4
Core heatup and melt release, and vaporization release (gap, melt and vaporization)
R S
Core heatup and melt release, oxidation release, and vaporization release (gap, melt, oxidation and vaporization)
Table 3.6-2 DISCRETE TIME' PERIOUS DEFINED TO MODEL THE VARYING EFFECTS OF CHANGES IN CONTAINMENT FAILURE TIMING T
1 From the . time of accident initiation to initiation of core overheating T
2 From the time of initiation of core overheating until the time of pressure vessel failure T
3' From the time of vessel failure until soon 'after vessel failure or when core-concrete interaction occurs T
4 Long after ' vessel failure or vaporization release has occurred -
v - - -,.y - - _ . - , , _ , y__, .. 3 __ .%_., - ,_.y.. . . , - . ,n- , - - , - . , - .
~
.I, ATTACHMENT 2 Conditional Probabilities for all Release Categories
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