ML20140E753

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Forwards Results of BNL Preliminary Review of Containment Response Analyses in PRA Performed by Science Applications, Inc.Highlights of Back End Review Results Listed.Comments Requested by 850601
ML20140E753
Person / Time
Site: Limerick, 05000000, Shoreham
Issue date: 05/21/1985
From: Bernero R
Office of Nuclear Reactor Regulation
To: Thompson H
Office of Nuclear Reactor Regulation
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FOIA-85-772 NUDOCS 8505290001
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'~g [3 g NUCLEAR REGULATORY COMMISSION ~ s {, 3 s.,(:. ; wAsmNGToN, D. C. 20555 \\Q*.C R+ / MAY 2 1 1985 MEMORANDUM FOR: Hugh L. Thompson, Jr., Director Division of Licensir.g Th'mi:s P. Spets, Director e Division of Safety Technology James P. Knight, Acting Director ~ Division of Engineering ~ f. FROM: Robert M. Bernero, Director Division of Systems Integration

SUBJECT:

PRELIMINARY REVIEW OF THE CONTAINMENT RESPONSE ANALYSES IN THE SHOREHAM PRA We have completed our preliminary review of the containment response analyses in the Shoreham PRA. The Shareham PRA study was performed by Science Applica-tions Inc. (SAI). The Reactor Systems Branch, with BNL technical assistance, has reviewed the containment failure and radionuclide release analysis. The results of the BNL preliminary review are enclosed. The Reliability And Risk Assessment Branch (RRAB) with technical assistance from BNL has reviewed the internal events analysis to core melt. Their preliminary results were submit-ted to the Division of Licensing on December 31, 1984. It may be noted that the BNL contract for the "back end" review is a liiInited one. BNL concentrated their review on comparisons to Limerick since BNL has gained extensive experience in the previous tr: view. The BNL review (Enclosure

1) has thus concentrated on areas where there are analytical dif'ferences between the two PRAs or differences in containment design.

The highlights of the "back end" review results to date are the following: 1. The dominant difference between Limerick and Shoreham estimated core melt frequencies is in the Class IV (see Enclosure 1 for definitions) ATWS with Shoreham being two orders of magnitude higher than Limerick. This is a risk dominant sequence because containment is predicted to fail prior to core melt. Much of the difference in ATWS frequency is due to the lack of an automatic, multiple train liquid poison injection system in Shoreham. The Bg/ reactor year.l estimate of the core melt frequency due to ATWS events is 6.10 The BNL review indicates that ATWS events contribute about 40% to the total frequency. RSB believes that increasing the liquid poison injection rate to conform to the ATWS rule requirements would reduce the ATWS. contribution to the total core melt frequency. CONTACT: G. Thomas, RSB x29445 4 [ l S60Wi" o l

6 2 EAY 211985 2. The release fractions for Shoreham are several orders of magnitude lower than both WASH-1400 and the Limerick PRA. Most of this difference can be attributed to high per.1 Decontamination Factors (DFs) assumed by LILCO. BNL used SPARC to assess the potential for lower decontamination factors for a range of conditions. The latest BNL decontamination factors are included as enclosure 2. These are time-averaged pool scrubbing factors based on SPARC (NUREG/CR-3317) { and do not take credit for in-vessel retention and primary containment hold-up. Note that for the Class-I sequences the suppr-ession pool is subcooled and the BNL decontamination factors are high and in substantial i agreement with the Shoreham PRA results. However, for Class-IV sequences, the pool is heated to saturation before core melt and the BNL decontamin-ation factors are much lower (600/100 vs 50/22). For the Class-IV ATWS ~ with failure at the basemat (y"), it is assumed that the pool is relocated to the annular region of the reactor building which surrounds the primary containment. Thus, the in-vessel, release through the SRV's see the same scrubbing as the ex-vessel release through the vent pipes. The slight difference in the decontamination factors (9 compared to 14) for the two releases depend on gas blowing rates and carrier gas composition at the time the scrubbing occurs. 3. The Shoreham PRA is ambiguous on the possible failure location for con-tainment overpressure failure. The location of failure is a key element for assessing the potential pool decontamination factors. The BNL tech-nical assistance contract has been revised to provide.an independent assessment of: (1) The containment failure location,and the containment failure modes. (2) The pool thermal / hydraulic response, including possible water relocation for failures in the wetwell region. (3) The pool thermal / hydraulic communication between the pedestal region and the outer annular region during debris / water interaction. (4) The likelihood of secondary containment failure and the potential for pool bypass given secondary containment failure. This work is expected to be completed by July 1985. 4. BNL is reviewing the impact on risk of the new source term methodology being developed by the NRC Accident Source Term Program Office (ASTPO). The reports issued by BCL.(BMI-2104) presently do not assess the release rates for Mark-II containments, but the accident types and containment performance are similar to the Mark-I results. In addition, BCL is currently performing release estimates for selected sequences in a Mark-II geometry. BNL will evaluate the impact of these results on the Shoreham PRA. We note that the draft report from the American Physical Society (APS) emphasizes the importance of containment performance and has taken i the position that much of the source term reduction has yet to be validated.

6 3 MAY 211985 5. In the preliminary review, BNL concurs with the Shoreham PRA release rate estimates for Class I and Class III accidents with. early core melt and delayed containment failure. For these classes of accidents, the suppres-sion pool remains subcooled and effecti,vely reduces the source ters. For the Class II loss-of-heat-sink transient _s (TW) the BNL audit calculation: using SPARC indicate reduced decontamination factors comgg/ry) and rela-red to the Shoreham PRA. The combination of high frequency (1.2x10 tively high release rates may result in a significant fraction of the risk for TW sequences in Shoreham. Note that the unusual geometry for the Shoreham containment (supp-ession pool vent pipes below the reactor vessel) negate the APS concerns for sustained core / concrete interaction and resultant vaporization release for ~these core melt classes. 6. Risk estimates for Class IV and Class V core melt sequences have the ~~" largest uncertainty. The preliminary BNL review indicates that the Class IV ATWS will have a much higher frequency (Item 1) and a lower decontam-ination factor (Item 2) than the estimates in the Shoreham PRA. The frequency of the Class V containm9nt bypass sequences have been increased in the BNL review (to about 4x10 /ry) and may contribute significantly to the Shoreham risk profile. We request that all feedback and comments from all NRR divisions and LILCO be forwarded to us by June 1,1985, to allow us sufficient time for consideration in our final evaluation. . contains the preliminary report from BNL. i k obert M. Bernero, Director Division of Systems Integration cc: w/o enclosure H. Denton 1 D. Eisenhut A. Thadani A. Buslick M L. G. Hulman R. Caruso E. Chow F. Eltawila M. Wohl 4

a .2 ~ '- ^ .i. ~~' ~ ' 4 i ENCLOSURE 1 Q BROOKHAVEN NATIONAL LABORATORY MEMORANDUM e oars: September 27, 1984 TO: W. T. r t K R. e ins and S. Y. eh-FROM: Preliminary Review of the Containment Response Analyses suexcT: 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 development and quantification. This " front end" evaluation is a much more extensive review than the present review and it has provided many valuable in-We have concentrated our review on comparisons to Limerick since BNL sights. has gained extensive experience in the previous review,2,3 and the plants Our review has thus concentrated on areas where there are are very similar. differences between the two PRAs or containment design di ffer-4 analyticalThe degraded core frequencies for the four accident classes are shown ences. 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. Cla'ss 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-In other respects the scram tem and to di.fferences in the A05 inhibit logic. systems used in Shoreham and Limerick are quite similar. t ,,,_,m,_,.,,,

s 2-September 27, 1984 W. T. Pratt 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 calcua 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. 7t will be in-portant to verify that these reduction factors can be achieired under all se-quence conditions and failure modes. DISCUSSION 4 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 (CgRgT -y), SNP-11 (CgRgT -y), and SNP-12 1 1 (CgRgT 1 Note that SNP-12 consists of both y;-y'; CgRgT -y") as described in Table 4. and y" scenarios where the y" scenario assumes the wetwell fail'ure.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 CFT for the various accident sequences are summarized in Table 6 with the implication that the DF for 1" 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 -y) sequence in which 10% pool bypass is assumed] before entering the i 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 c.an be made, i l 1. The diaphragm floor at elevation 62'6" of the Shoreham plant was not an-chored to the containment wall as in the Limerick design. The Shoreham containment wall displacement will expand outwardly under pressure as shown in Fig. 3. Based on this free standing diaphragm floot 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 ~-. ..,,,-.,n-----,..m..,.,--o,Ec.-- -en , - - ~.... , ~. -. - - -,

6, 6 ~ W. T. Pratt September 27, 1984 failure condition, the suppression pool integrity and high scrubbing ef-ficiency will probably be maintained. However, when gross containment failure occurs, as will be the case in many Class IV accidents, the base-mat-co'ntainment 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 thef 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 s not include the At present, the Shoreham PRA doe'.,Instead they are evaluated explicity. y" scenario in any of the sixteen release calculations. " binned".with the y' sequences where the pool remains intact. 3. After the bottom hea'd failure, the Shoreham PRA predicts that 90% of the core debris will flow to the suppression pool via the four downcaners 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-erste a substantial amount of gases to threaten. the containment integrity. The esticate of 90% of the core debris flowing into the pool is p' obably r 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 nore 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 fail ure." Generally the high temperature molten debris (~4300F) is takeq to be the limiting case.,For Shoreham, however, the low temper-ature solid debris (~2700F) nay be the worst case since very little de-bris would flow through the downcaners. 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. 4 ~~ := ~

W. T. Pratt September 27, 1984 4 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 pipes. 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 l 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 a'pparent that these releases have been binned " downward" into the lesser release cate-gory R,. 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-I quences D6 and D8. The completed table should be provided, i 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 i ranking. In particular: the use of I as the surrogate for melt release ignores the fact that there are noble gases in the melt release which i will not be scrubbed at all; the use of a large scrubbing factor (500) for C, transients is inappropriate since most of the melt release will be released directly to a failed containment; the reduction factor of 0.01 t for y" failures is indefensible since the event tree precludes everything but large ruptures where the pool will be blown out into the reactor i l l i i l / ~,, - - - -. - - - -.. -.

._,.,..n,_.n.,

) t 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 Jouble scrubbing with a decontamination factor of 50000 for an event in which the final level of water in th'e suppression pool is highly uncertain. Table 3.6-10 should be replaced by a table with defensible reduction fac-l tors. As a minimum the table should include a separate category for CS i 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 i full-size legible copy should be provided. RECOMMENDATIONS In addition to the above information, we feel that there are several areas which are important,enough to warrant independent verification. The ba-a 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 cr on the drywell fl oor. We propose to run a Class I accident sequence (e.g. TQUV) with 505 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 r.o quantitative analysis to preclude failure of the wetwell below the waterline. In fact, Appendix M indicates that the j 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-j 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 1 mitigate rel' eases 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 ass'umption of no pool-DF for this sequer.ce (n 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 i _,_-_._..__-,_,_m._. _ _. ._-.-.m,. -._..y~

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 explo.sions remains un-resolved, but the issue is being addressed by the RC as.part of the Con-tainment Loads Working Group (CLWG) effort. We propos'e to review the preliminary results of the CLWG effort to ensure that the MC positi h is consistent with the low probability of containment failure (4x10-'p) 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. Shorehaus Nuclear Power Station, SAI-372-83-PA-01, June 1983. 2. H. Ludewig, J.W. Yang, and W.T. Pratt, " Containment Failure Mode aind Fis-sion Product Release Analysis for the Limerick Generating Station: Base Case Assessment," BNL Informal Report, BNL-NUREG-33835, April 1984. f 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. "Reactoi Safety Study--An Assessment of Accident Risks in U.S. Comunercial 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 I)," Appendix D of the Containment Loads Working Group Standard Problem Results, to be published. ~ KRP:sm/tr cc: R.'A. Bari G. A. Greene O. 11 berg j H. Ludewig i K. Shiu J. W. Yang R. Youngblood W. S. Yu -, - - -,.. - ~ d.,--, ,-----w ---s-,,,--- ,,,---n,-- - - -, ,,-,~n-e. -n,.-------

.. = 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. d. Accident sequences are totally reevaluated using the latest BWR thermal hydraulic calculations for trans-ients, LOCAs and ATWS. e. Component failure rate data and common mode failures are reevaluated based upon the latest data from l 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' dasign. The probability of this, event is reduced. h. The conservative assumption that all potential core' damage. sequences lead to a major release was reassessed. Detailed containment event trees were developed to appropriately characterize the accident sequences which could lead,to a radionuclide release. .v., ,e, ,_-,_-_._m_.,~ v. y_ _.... ~ .,.__-._..-m._,

o ~ Table 2

SUMMARY

OF SIGNIFICANT RADIONUCLIDE INVENTORY RELEASE FRACTIONS BY RELEASE CATEGORY

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TABLE 3 Comparison of Release Fractions for ATWS Sequences Iodine or Cesiun Tellurium Shoreham Release WASH-1400 Limerick Shoreham WASH-1400 Limerick Shoreham Category 1.6x10-4 .3 .1.762 2.5x10-2 11 .11 .07.732 I Based on comparable BWR relea'se assuming elemental iodine. 2Depending on failure mode. e. e 0 J 1 e

Table 4 DESClilPTION OF Tile REl. EASE CATEGORIES IDENTIFIED FOR Tile Sil0REllAM PRA (Slicet 1 of 4) Y ettIau Onselsiast ACCINhi (lI MAR N$ttlPn04 SEQufhCE (estteltullen P90CM5510e fAtu CAltsamt BAltl fee le-Ptant anattill N 580 4 tot EIIggg*I Shts.3 This release category is representallee of (188s I acCldent sequences test Of effeslie Power feltlater fellere la.el. lag a transient e.ent leadlag to core meltde.n me.ere the con. to race.er wi.lsten i er ll electric pe=er. talament f alls to lietate or f alls hy e.erpressiere early la the secl. failure of high pressere tejectica systems. Early f altere in dent setwence leaJIag to 4 leslage type release free the drpell. failure of A01. fallere to Iselste la the the Sapell drpell. Ihis release cateto el.s reprettete,lles el Class I accident seemences Se fleed which portf ally drelas suppress. lee C 8,fi.3 ShP5 2 ry i ..ei.i.,a tr..s to.here t e ce. tai.ne <. n s i. isei.

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s. 3: -

s i: es.-!sg:5 J:_,: 3 = c. =m. es : .=2 =3 _. : i 2 g-i u. =g o _ a.f: 2 2

  • -= 5 3

z. 4. . : r, .2 ..w _=g,- 3-3:4{1 4 s a.:! g:33 -s-33 i L:xs: L:21-

1J s E--s 3.

mze,:

_31 3_y
g.,

..1 1 a

3. _3=g.

1 v

_3._

4 m .e. .:_ s : .:_ = g

..s

_23.: 2. _2 s a _. Uk e 4 I 13 4 e E A l Ih I k I i l 0 t I l - - -.. ~ ,--,,----,---,_m. ,e---- -.-----v..,

_ =. _ -

s Tabic' 1 DESCRIPTION Of Tile REl. EASE CATEGORIES IDENTiflED FOR Tile Sil0REllAM PRA (Sheet 3 of 4)

MiIatt ethlmAst atC:Mht (tI GimieAL M SC. PIsole Siquinct Cahiessulies ,seCMilies, Aim (alawai eallt f0s it. plant anAttlil M 5 r.1Alos 1 nu..i.u. c.i.,,, e..i . ci.u ::: uce=. www. a,*i= toca,e.n,..,ni,n.w w,,n w,. c s v.v aia in c ui -. e.n i. n. io,i... i.wi.e.. i..i. sw s., 3 s.hcisa niu=i. su <= usa. u Nut c 's,*

  • 3

,um e, in. -i= n. ro. w, u = i. caesi =.t t.o. is de t.S n. 'diaw"* a saw ** rid .= 3 ik tal. ..,.ua i.,aw. n..e,m==,ai.c n=.nu un i. iu, g.,,,, n,,

s. ca us =.

,,s., a ui i,.a. a en.. ir ,, u, e,,,,n i nei r.inu c..,,, a r.,r u.uu. er erwis en.ts misca..sivs.tenw..e ut. un. c,e,r., snes.ie .:.ii s. I i. i.e., ci.a i. ti.a n...a cini se.ui*.i ww=n aw. =. s.n.su u i,u v.i w i.,. ui..

n. i m u usa,=i si,u i n n,n,,n sa..a a. c a n i a.

uinio e,= iwieill.r.gui ,,w is.. _in.e i'. i i.ncuss. '"'s'd,"* i.e.,,n, a ini.ui, i. a. ucia.===.. ,wiu si,u n,se i= e,w n b,,.us.e '**8"" in. i ,n u i. sws.u una r.i.u.c.upr, m.i...ca.n av use. w w =.in mov u. . a=s t.nw..t ac, e.siw. - c,,,8., asa n. cous e.ini n, t.n-. i. u, .w r *H 1.hasa irisas hlws., causa- =*hH-*.** Hida,<.*cu aiand 4

  • c., n.a. s.n. n, cu.

iin. in. cani. a.isw. ,,,,',,n"' 'p"r.!"i.e",'"i."."'i* ' "* "" " 'a""*i-si u w.a u.u-i. a. u, n.a sw_ nut.u.r.4=ts u.un. ia.<i-a asi.a.e i w aam..e i" '" "'"i i "!- La -** ' 1

,i,.i,iim. o ii istiv cl.im,e Aful, f.Ilw..f StC. f.Ilare c,e,f,.1' Sars.it insi.e. lees. seiegory is i.ee a Class if sesideal lequeste le of all lajectien systems follows.g sentana.

gag 448,9 .hich the s.mtalasset falls by a fallere le wres and reme.e most faHere, the fisglen preenges released daar b.el, fellemed by core seitdeun. the c.ateiament tellere fr the feel 0 q tggl, p g,gelg y, $re.g. g,,g' g,g is assesed to occw le the met ll and the fhiles products are perted to lhe,,sostal.a.eg,jgggg_lp'sep. slyelligently aHenealed prier le lit leelete le the emetteensel, pt9Hlei~pieI.~ f ] s I

u. Table 4 DESCRIPTION OF Tile REl. EASE CATEGORIES IDENTIFIED FOR Tile Sil0REllAM PRA (Sheet 4,0f 4) i 4 e 30MinAbf ACCIN NI (tt ett8a5t EnteAL M Maltiles 544utnCE ( Mieltuiles PeetM551on FAfit taltiast 34515 70s lu PLAh! ANALV585 MllCsafet laterfulag 18CA. the supprestles peel is gggg f Ser5-13 Ihis release category is represeeeettee of Class V eccideeg partially effectlee in altig4tlet releases. 148 scenesces which levelee core meltJe=a fellewleg a (MA out. i slJe centelament. The sevs are acteated in order to altigate the rejsess of fistlee products to the envitesseet by providleg se alternettee path late the castelement (f.e., supp.e stes peel) durleg the la.eessel release perled. J i 1 Shr5-14 Ihis release (etegory Is representatlee of Class V accident laterfeclog L K A, fellere of Says. gag g 14I setusace =hich levelee core pelade-a felle leg a tKA out. tide costatement. the Sets era essened est to be opened, sad the fisslee preewcts selefsed free the feel tetelly bypess the gestelaineet, t Shel-Il Ibis release cetagery is represeata' lee of the teentested core toss of caedesser secuum, failure of blgh gg9.g neltas e accident sessences la which the castelement remoles latect pressere lejectles systems, fellere of Asle 454 end the release of redteauclides to the emeltesteet would be very cestelmeest telegrity is esfA aleed, small, med deteraleed by ledese to the router tulldlag. ) Shr5 36 This releese setegory is representattue of the tesotested sore test of sendesser escomme fellere of hien C e T *I seitde e usedest seg eases 8. =hech sne sostelement ree las setal pressw. s#Jailes s,stesse fellwe et ms 154 e and the relegte of re lides le the emelreement heuld be eery 6804 element lelogrily is selsteleed. smalle and deteseleod lee 6ege le the teatter belldleg. forehere sere, the releasel usuld be filtered by the steadby filter systems. 4 ee -.- Y 4

O Table f SHOREHAM

SUMMARY

OF DEGRADED <0RE ACCIDENT ANALYSIS CASES CartfA!x. tnf PR!PARY SUPett35fCN POOL F L ! St1 TIM CtmmluTM CWC3 IISCO (c) CIII W 3 CC3tC:iArca(s) CCPCSITICle(h) FACTCA5 Say vtNTE 34g.4 CI W T M 100 N.C C T CtTg4g-3 Ct W T 120(e) to Cd3 N.C I C t 7,.7 p-W T M M N.C g4 Ca7g 4 4 7* p.W T M M N.C N.C C '4 g +7 07.W T SCO M T e 2 C8T2 4 g. 7' W.W 7 M M N.C 100

  • N,C C2T3 4 g.7 CP W L

100 N.C C27344 7 07.W L 100 N.C C27344 7' CP-W L t CtTM pW T(e) M M N.C a g C2Tl7 CP-W T 600 100 N.C 44g 4 4 g. 7 C2T QP-w T ese les N.C t 4 cat.A SP t(f) 2000 100 N.C.R g4g i CsTg4g.e 6 L m M.CJ, i (e) CI. Containment Isolation fatture 07. Contairrent Overtressure 14tivre Cd. Cantalement tresca is ta the cry. ell (7) w. Cantanarent tresca is ta tre wet ell (7'l (t) T. Transient event ita srwery syste9 tatatt aae en effective reteation of 801. L. LCCA event esta tae ef fettlee scimary systee retention of 23 and IC for tae vasors one seresels rescettively. a (c) =. PM C M C. C*AA/11 4. Cante-st wsvle of U*tP.! (e) 80el samtsing ef fects.eaess is reewcee eve ta tae resuced suneertence meset. (e) fan setteat af tre f.ssion aroevets releasee 'e-A tre c;re regien witnia tre tr*.ary siste9 tysasses ta.e seel. til Fif ty ;ercoat of tee fission artegets releases frcre t?e fwel is etrettee late t*e c:ataire.at of asening all tee 1Af sistMarge Itaes cartag care Peat va and teitco a.

s..' Table 6. SHOREHAM POOL SCRUBBING DECONTAMINATION FACTORS PATHWAY EVENT SRV DOWNCOMERS Class 1, 2, 4 3000 100 Class 3 NA 1000*, 100 I e 9 O e e9 9 e e O

.i

  • 22 l
=

g 4 3 S S 3 3 3 s: 4 e e e EEI E sv: = -32 9 9 9 W-rgg i xl I a e e e a e e e k-Ij l a e e e a e 4 4 I' ' -5 I Ia e e m5 L.) ! I - {- .e e e. e e e. e. e s -I* g [=u E"1 G E I s s s s a a

c..

g .a m I w e s s s s a a v.c a. e e e a e e e ,F o O o E i I. ij 1 8 8 8 95 2 ,= s.j3:. m m ,s 1 2., 9 ". r 2 9 a 3.3 .= e >- p 1 U1 4 gg j

3.. -

as = 9 m 4 .I w E v.: a-3 YK 6_. -l L M M s e 2 II W y I I

1. *332 v

9 3 II -t i i m x !JI C ~T N 3gE' s S 2 '\\. a e a e g s-l >$ I.I] I. .1 l-2*E m

2. jl's 18 1

I s

s 's 4

e a .g z a I 't g2 1 .-{- 2 5 g_ 1.: 3-i a 4 15 3: a = a: 4e

=. - ga s r: i as a

4. -= i =. [ .= 1,. .t l .,1 -= 2 4-lI' y.-j i r ! I h.8*1 4

t y: F i

=3.: 2 o 7.... ~ g M - ~ ~ :- l 552 33 5 i 3 -.' 20 i ,, - -. _. ~.,___.

0 10 ATE FREQUENCY _.;_Sh5MS_2d.'_______EI5_'__"1._______ l UMERICK TOTAL io-s ' ~ I l I g l l l I l l l l I l' I k l I l l [ddyrhWg 5 30 ,,_2 4 10 10 CLASSI CLASS 11 CLASS lit

  • CLASSIV JT MOVA

'"?,S'u',"' l ^J'"" CONTAINMENT Figure [ Summary of the accident sequence frequencies leading to degraded core conditions summed over all accident sequences within a class.

N.htNEc$3'YsIt!$t7' r ceYEi$e'ldYtaIesf 10-3 8 5 ano wt in? 4 3 2 p p ,4 -4 i n v p ppp swongsan. wAsu-14co // p p p p y a va$^{gg ^" ' I '" ' ) va$!!sts ^' 'l gG'& conNar va$$$stt WLh!!!ett cenNe.r "'^^"I m c:nUni cc2Nnt ~ C0"E "'T PUBLISHED B'.4R PRA RESULTS 8 .e --w. --,-,r r ---we,ei, = -,

w---~--y-------r v--v---w--- - - - - -, - - - - - + - - - - -

-r-- - - - - -. + - -r e-v"*e*

Ili C ~ ~ ~ ~ ' Vertical Disp 1 Radial Disp 1 DB o.g e I I I i I t 0 j l l ~ k l t I \\ i g s/ s 'g G M 99I s. o

  • o7' i

/ o e / I / / Y d ,f Y,, {} { Qp = go s 4 9 .-.---,---,--.rw

l 4 .yr l ~ l 4 4 l l 1* GROUT sitat iians azu RUCTOR SUPPORT fiq r C3CRCt m y SMAA AING 13* 4 g A ?< i*.2...,,W f* 24 32 3/4* EL. 63'8* ~ AiA M T QPIMi;tr. l 4 } f l z;;..: nv - m i P'

  • 1
u. w r I

s t n.acau. '. i' l j r-I l l t I l I A ? * -'EOS 00WIC0fetR i Figure i Reactor Pedestal Schematic Showing Reactor Downcomer Gecmetr) ~ i I 4 =+-m --r w m-- ,..,w- +---------g 9 g - + - y-w---n-

ATTACHMENT 1 Definitions Used in the Shoreham PRA Accident Sequence Classification 4 4 e O e o I e O 4 e e I e

l-Table 3. 3-1 '~ ACCIllENT SEQUENCE Cl.ASSIFICATIONS USED IN THE SYSTEMS EVAL.UATION M(Ind et slosta t PNY11(AL 8A115 511tist tf f(L p(PetighfAllgt piticar:fus fit (4 AlltfICAlles (0nitituilleG tytti SiquillCE

  • SEQuthCE fee CtA15 Class I (C )

Selatively replJ core selts centalament Transleets involulag less of leventory easeups - Iransient with less of g Intact at cose melt and at teltlelly ordlem and small (SCA events levelvlag $89 octs. high and les presswe j, Ise pregsures levelres a release path. ellen with less of laventory easeup; traestents coelaat esteep v4y fees the vessel to the supprestles Involvlag less of scram fonctlee and lastility geel to provide suf ficleet coolaat seseep t '#ii'*t# ' i"' '' ** i t d*' Class il (C I met heat removals leadverten't SAW epeali" ' *t *t'i"' t'**'i'*t "iih i *t T##**$'*t' *# ieCA' i"'*t'i* 2 accl resideal I. eat removal J q le=er decay heat powers contalement is failed refer to care sells levelves e dents with leedryste best reeeval capahl Ity selease pothway frue the vessel to the suppsesslen peel ..a e i U taats lit (C l seletively rapid cose molta cantale. large 18CAs with lasofficleet coelaat metreps large 10CA with less of j . 8.tect at core melt but at small and medius IOCAs with fellere of the Sets les pressere f(C1 satttally bish latesmal peesseres le actuate end I.eg teen less of leveetery involves a sel.ase isee the vessel massep3 SPW failures with lesofficleat coeleet j' to te.e daywell

e. teep I

liess IV (C ) selativelp repl4 toes solta castele. Iseesleets leveleleg less of scree functies eel Transleet with istlere g t falls prior te cose eelt due la less of castelament heet re-eval er all veutle. of hP1 and is18ere of .a, reessese; levolves a release path. ever Ity ca srels treestents with less of scree fenc. St(1 f..e the vessel to the s.epresslee Blee felle-ed by acteated depressorisettee j Puel ~ i i 1 1 Class V lC ) selalleely replJ care esill teelsle. 10CAs selside centalement with lesofficleet itCA le sete stees g failed free Initlettee of accl. seeleet easeen to cores laterfeclog system LOCAs ilmes =lth fellere of 3 i dret ese le evelleest falleeeg imselves with lesesfficleet toelaat esteep 3151f closure end less j a selease poth ar free the vessel seith ofICCS 6, passes the coalelement M I e I i s e

Table 3.6-9

SUMMARY

OF CET CATEGORIZATION l i DESIGNATOR DESCRIPTION ATTRIBUTE GENERIC ACCIDENT SEQUENCE CLASS C Class I Pool scrubbing prior to g transport in contain-ment. C Class II Pool scrubiiing prior to 2 transport in contain-i ment C 3-Class III Pool scrubbing is by-passed prior to transport in contain-ment C Class IV Pool scrubbing prior to 4 transport in contain-ment C Class V S Pool scrubbing and containment are bypassed RELEASE TYPES R Gap Release Recovered accident after g ~ initial core overheat R 'E3 0'P '"d H'It D1"' R ''d Id'"* 'f**" 2 oxidation release significant core melting R,R5 Gao, melt, and Unrecovered meltdown 4 vaporization accident release plus oxidation release s 3 417 l.

Table 3.6-9

SUMMARY

OFCETCATEGORIZATION DESIGNATOR DESCRIPTION ATTRIBUTE GENERIC ACCIDENT SEQUENCE CLASS C Class ! Pool scrubbing prior to g transport in contain-ment C Class II Pool scrubbing prior to 2 transport in contain-ment C Class III Pool scrubbing is by-3 passed prior to transport in contain-ment C Clas.s IV Pool scrubbing prior to 4 transport in contain- . ment C 5 Class V Pool scrubbing and containment are bypassed RELEASE TYPES R Gap Release Recovered' accident after t initial coca overheat 1 R 'E Gap and Melt plus Rec vered acciden: after 2 3 oxidation releasa significant core melting a,R, Gao, melt, and Unrecovered meltdown g vacorization accident release p1'us oxidation release e 3 217 ~

~ Table 3.6-9 (Continued)

SUMMARY

OF CET CATEGORIZATION DESIGNATOR DESCRIPTION ATTRIBUTE LUNiALNMLN4 FAILURE TIME e T Time phase T Containment is failed 3 g before core degradatic'n T Time phase T Containment fails during 2 2 core meltdown in vessel T Time phase T3 Containment fails 3 during cofe-concrete interaction T Time phase T Containment fails in 4 4 the long term CONTAINMENT FAILURE MODE BP, CI Containment Bypass Containment isolation or Isolation Failure fails or containment is bypassed 6 Leakage in the Dry-Leakage sufficient to. well preclude overpressure failure l 8 Leakage in the Wet-Leakage sufficient to well preclude overpressure failure OP Overpressuri:ation 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 4 1 3 218 l ~ " '

Table 3.6-1 TYPES OF POTENTIAL RELEASE FROM FUEL ^ NR No release R Core heatup (gap) y R Core heatup and melt release (gap and melt) 2 R Core heatup and melt release with potential for oxidation 3 release (gap,. melt and oxidation) R Core heatup and melt release, and vaporization release 4 (gap, melt and vaporization) c, R Core heatup and melt

release, oxidation
release, and S

vaporization release (gap, melt, oxidation and vaporization) Table 3.6-2 DI'SCRETE TIME PERIODS DEFINED TO M00Ei. THE VARYING EFFECTS OF CHANG IN CONTAINMENT FAILURE TIMING i T From the time of accident initiation to initiation of core g overheating .T From the ' time of initiation of core overheating until the time of 2 pressure vessel failure T From the time of. vessel failure until soon after vessel failure 3 or when core-concrete interaction occurs t .i T Long aften vessel failure or vaporization release has occurred 4 i b i_ -..._ _ _ -. _-,_ _ _ - _ - - _ _. _ - ~ _ _ -

ATTACHMENT 2 Conditional Probabilities for all Release Categories 9 e 9 a= 6 9e 4 e O 9 a E e

=w ? E.s = ???? ???? ???? ???- )sE;33 u=== wwww

==

=w

ig=55 dddd ""dd dd ": 4d - d-t tj-1; z h yy i

II:=e

.~ l* i u !t*' yj'c inaia agy - = t g [ $d.k..* =M M = su s M ???? ???? ?? ln= n M E 8

.! * * * * *

  • n!

4 ,i m g i 4g ua ,g ei t 77?? ? ? ? *t' ?? I = em D 3 = 3 2. w. *.M. x. =. 3 3 m. is a s y

=

l. =. uss l ,g g m A w a w

o. o.

? a. m E 5' MM M8 .: 4 d4 W E ???? ???* ???? ???? ?? TT a

==

==

==

wwwm

== == m 4 4 a.a 4444 4444 addi 44 tl ai d . w C f ???? ???? ???? ???? ?T TT E Q mzxx zum= warx xxwx 33 zu 44aa 4444 a44a J4a4 4: 44 p.- cc :c ED ?? l? ? = a

== l = a= 44 e .s. f = .s m g m I. l D ." 3 9e = a s !s s v 44 l 44 a ou. I E ?? t? ? d 4m 3m 4a j d.3 i p w {

e. e.
o. e.

\\ as 3 M= .WW j 1 de as _ e. a M*33 w3** N3 2W i 3 ~ .==. asaa 44a4 44 d ei I l e.- ? ??*? ???? ??

  • ?

i G 3 = MWWW WWX* WE WM ) a adda 44aa 44 aa v W 88 = 8 2 2 M tt C 2 I=T ii.L. L L L. i a. a.. & L & i_. a a a 1=L A& aa&& a, &. L' = = - ----

=

.==

=

.l e e w= 1 3 a il 3 t 1 H.63 i ,,,---,m,--- m,,- ,,.,,---,~_.,,,,,.,.,,,,,.--.--,,,-wwn,,,-----e..-wnr. -,,-n--,-,,-,e,-_---,,,-m,-,--,s,-r,.--~-n-,---w-,- - - -. - -, - -,,

4 t _l g3I ????.???ivvv????? I g*E 3

  • *WW WWWW W W W WIN *
  • 3

!3 ,Lggs. a easw

e a
4 : sessea

. :_ f 1 lj:2: 13 c 2, shgi fa:Ja I gai E t l e .5 e i m i l W l ] a a e _. _ g_ 5 a I .W m W 'I f W E I l m e W e I I l M l W 8 g -. I. g ...g... 1 <= 3 ,T E a A E ~ o-e 1.. .2 m 4 m 3 a n I v g l l -W a t a J g I s ) 1 d '8 i E \\ i G e 8, a l' y l1 ) l l l l e i 8 1 i i s. j f I 2;. 2.. I, i g c:' i .: i 2* I* ; E ;h 3 - h h

  • 7 i b h," h I I
  • I I I C h h I N ~ ~

~ l h., o 3 g l-al. E i _2 11 f:a ---,,,e-, ,_,,--.-.m.-- .w-,.--a, ,,,...,---.w...-- ..,y- <, - -.. w e, r-w- + +. - - = w rw-

  • -r

--we. w-. - - - - -&

e 4.l 4 3F: .}-al. 5! ???? ???? ???* T??? www= w :1 = = ww=mt===wi ~ 3 44a4 4.:44 444484444l g i i 5= an[=[ ?? ? 3 ,=w ca=ja ww g: "~ ?? *??

n
u. n.

u n. = t - a. ) r ?? *?? m =. a. a. a. =. =. =.. p ua 3

      • =

Nw** .i M 4 l.9 8 .: 44; 4424 8 w.. M l 0 ?"?" ???? f ?? I? ? W 3 22W= Ww** l WW ;W w 244; 44aa jd 4 4d W w .J s s.e s.e es.o y WW WW 4 a g 44 dJ u W l TTTT a"TT ???? ? " *l' ? ?? !? ? 3

==

wwww==== wwww WW iW W M 4 4 4 a. 4444 ass 4 44as 44 [' d. + g 5 ???? "??? -l' T T T "??? ?" i ?? E E

    • ww wwww **ww wwww Ww l*a d W

o d.; 4 4 4444 age 4 4444 de e- - 45

a

?? IT T s WN 'MW 4J I44 as ,..a_. l ? '" .m M I I W WW f3N .d l 44 14 4 5 ?* ?? l4 4 I NW WW g" 4 .; 4 g p.= s $ 5 le e ee l$ $ ~~ s 3 4a 44 Ii ?? lT T a. g S A$$$ $ES$ I MI I48 e 44aa aadd f ~~ ? ???* ???? ?.T !?

  • l4 4 8

g .2... MNNN NNNN WW =W 4 44;4

444 44 l

u 4 i i i, 5 i 5 2 I W 3 .e .t =

2. : 1: : ;* 2,2 :

l Ehg h *

  • w b Y b. ; $ $ ;*

1 1 ; 1 4 2 C C ' 1 " E ' 2 2 22EE lt

== l 3 fa 3 5 3R H-65 I f y-, z._--. ,y..- .. ~. _ -,, ..,------_-,r-m-mm_--. 2.-.

] .,2.* e* I i i r. EI. .2 ~ Ef t t ^ ~ ^ 2: 4 s ft_5:= 0: T r t.-1 r_L: i gg ? 35j35 Egg y 22:ix 5' .e. 0. u. 1 w a,. m = l3 44 44 m w .a ....w I

  • w "w w I

w x. x..w

z. z. _. x.

8 l-. W 5 ~ l 1. w a wxxw wwww w u in n m< J44d J444 l J4 f44 w .a

s..

w g sw as C 3

  • ., ed de w

>= ... ~. m g axxx rume ww ww w JAdi J444

  • 4 as e

o ???? ???? ?? ?? a w x. x. x.

x. w. a. x.
x. m.
a. s

~ E .w a C. = te ?? n= x

n. u.

u n. .m e 'i a m I'i t?.i ? .~ g l*M. ce C 6 m y a , M..M. .O .m ~. ~. n. C I

x. x.
x. v

~~ a c. t a ~. ~. 5 3 wxx azzz um xx 5 sees egs; 4; e4 -.g 3 4... v - ~ gi \\ s _5

t i

If : :.,:. l k! ! I

1. : : : :

,_.:_ __i.~ =_. i m o w w z = w w m m a = _ = = = = 3 = =,s = ig t i. .l l l' d ll E. 'I g't Y

    • 3 1

.a 25 4.j( _m ,_.-t

i i. 1 , s. l 3 "*- ???? ???? Ii wwawwaaw ng aaae eaae l', It Im. Is + n m E w i g P 3 w ~c w n W l W W a g i w i a: 1 ~ I w m I w w al dg t s e w .i El -o I 4 ew t I ~ 3 a 1 tv i t? 3 a! q ls u ,- ~ .~ it n ; w w = = =.= = w i4 ne aa j ??Ye ye?? != = ' !s s I.., 8 i .3 =5

    • =*

sz a-.-

e. 44

,.,g l 1 e ! l _1-11 'l =< i .t 3 2 u s j- .1-is-hik.z in i;; iin =iti.s it h sh i 5 5 th H i it h j3- \\ -g II Iy I I !!-73 x O e -m.~- ,-,-,,,,----,w--,.myr- --.,y------.,m.%---.----_.-..<.,-,.- ..~_,..r w,_ _.,-, w_. ,.-.,~w--erm-- v~'* = --'-'-----~--'*----'-v*+

z-C l !=i L1 ? ? '**% ' u ' = ;: ' x, s. s. s y!8!. .s.e ,4 a-e a, .n IIi.E. t s1I EI I1.333_3 .3. g_i s 4 4 232-J: s; i 1 .' 3 I y 9 i .m u m 4 4 w ~ 1 I h f N N N n a ,r 's .l ' i. c s. ' i. ' i. m m .s es w l 8'* t M J g j w = ~ ~ t Q 3 I

  • s'( 'ses

.2 ..s w w I 4 4 a t = w r g M ? i ? n d

  • = la w N, I 4 ! 4 W

m i ? ? ? ? ? ? ? 3

  • w'w sa v

'w'* ' m? s l y = e .s a a 4 a 4 a a g g g I e ? ? ? ? ? ? ? ? = T" 3 .w.* . u.w .u.w .x.w a Q s*es a a e a a a a a i t I a a M* \\ s l .= a ..T ..,? m - n l m ~... _ ~.. Q 'I a ., I .? !

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ENCLOSURE 2 REQUEST FOR INFORMATION a 1.'. Table II of Appendix M gives different pressure limits for the longitudinal reinforcement bars at the base of the containment and in the wetwell region. However, the longitudinal bars appear to be continuous and should therefore have the same stress. Please explain the basis for the different results. ~; 2. Table II of Appendix' M indicates that the shear bars at the base and drywell head have the lowest pressure holding capability (121 psi and 120 pii, respectively) but the discussion indicates that the additional reinforcement will preclude this failure mode. Since the containment failure mode is a key ingredient of the release estimates'7*please provide a quantitiative estimate of the additional shear strength provided by the non-shear reinforcement bars. 4 b o I i a a

3. If shear failure is precluded as discussed in Section 3.2 of Appendix M, "it appears that the ultimate capacity is controlled by the yield of the longitudinal and the hoop bars at about 123 psi." These two failure modes appear to be very important to subsequent fission product release (particularly for Class IV ATWS) since they will both occur in the wetwell region. Please provide an estimate of the size, location and direction. (vertical or horizontal) containment failures for each of the three i possible failure modes. 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 l structural analysis by Stone and Webster (Appendix M) did not identify any significant source of leakage. The basis for 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 phenomenological, basis for the estimate of only 10% bypass should be provided. Preliminary results from IDCOR 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 SR'V's and be released into the drywell. 2 L

-p \\ 6. The basis for binning into release categories is poorly described and the transfer from Tables H.4-8 etc. into the If. release categories is difficult to interpret. A table listing the specific sequences which are binned into each category should be provided. 7. The lack of R sequences in the release categorles makes it apparent that 5 these releases have been binned " downward" into the lesser release category R. The basis for this " downward" binning and any other 4 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 sequences D6 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 ar,bitrary yet it greatly influences the importance ranking. In particular: the use of I as the surrogate 1or 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 C 4 transients is inappropriate since most of the melt release vill 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 ^ building at high pressures. 3

O i Table 3.6-10 should be replaced by a table with defensible reduction factors. As a minimum the taole should include a separate category for C4 4 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 6 j the ranking process. This revised table will provide the basis for our independent importance ranking based on revised estimates of accident frequency and reduction factors. 10. Sheet 1 of Figure H.4.2 has been reduced so that it is illegible. A full-size legible copy should be provided. 11. Appendix L provides a detailed discussion of the disposition of the corium f f (90% is expected to go down the vent pipes) based on the revised re, actor pedestal geometry illustrated in Figure L.3-2. However, this figure is inconsistent with other descriptions of the geometry (e.g., Figure 2.3-2) .[ and provides inadequate infnrmation for an independent assessment of the } } corium disposition. Please provide detailed (as built) drawings of the vent pipes and their covers within and external to the reactor pedestal region. Include a description of whether the air ducts"and manways in the reactor support wall will be blocked during operation. 12. Provide the estimate of the fraction of the molten corium'which is expected to spread out of the pedestal area through the open manways and air ducts in the reactor support wall.. 4 i .}}

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