ML20212P264
| ML20212P264 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/12/1987 |
| From: | Sherwin Turk NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | Harbour J, Hoyt H, Linenberger G Atomic Safety and Licensing Board Panel |
| References | |
| CON-#187-2787 OL, NUDOCS 8703160049 | |
| Download: ML20212P264 (250) | |
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UNITED STATES , f ',f. . ./.J
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....+ March 12, (987 y WR 12 P3:38 Helen Hoyt, Esq. , Chairman Mr. Gustave A. Linenberger, Jr.
Administrative Judge Atomic Safety and Licensing Board Administrative Judg8FFICE Atomic Safety and L986his! 'OF,jtt{f[$y'- .
U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory mission Washington, DC 20555 Washington, DC 20555 Dr. Jerry Harbour Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, DC 20555 In the Matter of PUBLIC SERVICE COMPANY OF NEW IIAMPSHIRE, et al. -
(Seabrook Station, Units 1 and 2)
Docket Nos. 50-443, 50-444 Off-Site Emergency Planning
Dear Administrative Judges:
Enclosed are the following documents related to Applicants' 10 C.F.R.
I 2.758 petition for waiver, in accordance with the Staff's commitment in "NRC Staff Initial Response to Applicants' Petition for Waiver," dated - ,
February 27, 1987 (at 2-3):
- 1) " Review Comments on Seabrook Station Steam Generator Tube Response During Severe Accidents and Related Sections of Technical Evaluation of the EPZ Sensitivity Study for Seabrook,"
by T. G. Theofanous, dated January 12, 1987;
- 2) Memorandum from Charles E. Rossi to Vincent S. Noonan, dated February 9, 1987, with attached draft evaluation concerning "Scabrook Emergency Planning Study--Treatment of Preedsting Leaks in Containment";
- 3) Memorandum from Warren Lyon to Charles E. Rossi, dated March 3, 1987, concerning " Steam Generator Tube Rupture During Severe Accidents at Seabrook Station"; and
- 4) Brookhaven National Laboratory's final report, " Technical Evaluation l
of the EPZ Sensitivity Study for Seabrook," dated March 1987.
l Sincerely,
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i Sherwin E. Turk
! Senior Supervisory Trial l Attorney l
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Enclosure:
Service List M03160049 DR 370312 ' ' $07 O ADOCK 05000443 PDR
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i REVIEW COABIENTS on SEABROOK STATION STEAM GENERATOR TUBE RESPONSE DURING SEVERE ACCIDENTS (a draft NUREG report dated 12/15/86) .
- and related sections of TECHNICAL EVALUATION OF THE EPZ SENSITISTTY STUDY FOR SEABROOK (a draft BNL report dated 12/5/86) by T.G. Theofanous Depamnent of Chemical and Nuclear Engineering University of California Santa Barbara,CA 93106 January 12,1987 l
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- 1. Introduction It has been known for several years now that core melt accidents at high primary system ,'
pressure give rise to certain unique phenomenological suspects with potentially significant consequences on containment behavior. The basic phenomena involve strong natural circulation currents, of steam, within the primary system and if the system remains pressurized, eventually forceful melt ejection into the reactor cavity. A summary of the early history and my views on a number of related issues may be found in Attachments 1 to 4. A copy of some of the viewgraphs I utilized in the first discussion to claim that hich nressure scenarios cannot nersist to core melt is provided in Attachment 5. This presentation and.a -
subsequent one given in front of the ACRS (see references in Attachment ) contain the first suggestion of the need to depressurize the primary system once the severe accident progressed closely to the point of no return. A recent letter by Dana Powers (of Sandia) providing some important source term perspectives on this issue of depressurization is given in Attachment 6.
Finally, a copy of an early paper of ours which is referenced in Attachment I and which provides a scoping out of the effects of natural circulation is included as Attachment 7.
In order to complete the perspective created by the above,I would add the following:
(a) The potentially important effects of natural circulation were, probably, first realized by i
Vern Denny, who embarked upon a code development program while everybody else was developing 1-D forced convection models. The first published reference of this code development effort and some preliminary calculations of a general nature were presented in the September 1983 meeting at Cambridge, MA. This paper concentrated on possible effects on
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I hydrogen production only--i.e., no mention of steam generator tube ruptures, possible impact on avoiding high pressure melt ejection (and the Direct Containment Heating problem), etc.
(b) The first reference to the potential for steam generator tube rupture was given by L.
Winters in an ENC memorandum in July,1982. This was based on RELAPS calculations which artificially cleared up the loop seals allowing for circulation around the loops.
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N9- , a I t-3 (c) When we tied-in the natural circulation phenomena to the Direct Heating.and the Tube Rupture problems in the March 1984 Containment loads Working Group meedng there was _-
considerable skepticism (perhaps even some hostile reaction) by most panicipants. A whole host of subsequent studies, as well as the May 1984 meeting reported by W. Lyon and given as Reference 8 in the NUREG report under review here, were cromoted by this nresentation.
At this point there is mounting concensus that the system will fail, somewhere. prior to core slump in the lower plenum. But this is not a firm conclusion nor is it known whers the system will fail.
(d) Early on, we pushed for a definitive position to be taken (and documented) on the conditions under which the loop seals will cease to exist somedme into such an accident. This has not been done yet. On the other hand, the model experiments in Westinghouse (it must also be mentioned here that some scaling consideration for such experiments were provided by D. Squarer in the 1983 San Francisco Winter ANS meeting) demonstrated that natural circulation can extend well into the steam generator tubes even when loop seals are intact. The quantitative aspects of this problem are not fully well understood yet; that is, important modelling and italing question remain.
For over two years now my efforts to " move" the issue of depressurization in a generic fashion have been frustrated. As you will note from my recent, generic, letters (Attachments 1 to 4) I believe this is a matter of the utmost urgency. It may well be that a specific licensing action (such as the Seabrook) provides a better forum for coming to full grips with the issue.
I have serious reservations about the NUREG report both on matters of perspective and philosophy as well as on matters of implementation. I provided my own perspective first, because I think I can be more effective in my discussion of the report ifit is clear where I am coming from. But first let us address the BNL repon.
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- 2. Comments on the BNL report .
The relevant portions .of the BNL report include section 2.3, part of section 2.5 (pp.27,-28), and section 6.4.
Section 23 deals with the quantification aspects. The frequency of high pressure sequences in which SGTR "might have an effect" is taken as 4x10-5 per reactor year. 'Ihis number was estimated by the applicant, was considered reasonable by the NRC, and was used in the BNL study. I have no documentation on the basis for this number, but I note the absence of an estimate ofits uncenainty range. The NUREG report mentioned that the applicant meant it to be bounding. Without it this number is not very useful. In any case,it appears that such sequences represent a significant portion (~ 20%) of all core melt events.
Conditional probabilities for Steam Generator Tube Rupture were based on SAND 86-0119 addressing the Surry Plant, which in tum, were based on "expertjudgement," and were found
" consistent" with an earlier NRC memo (p. 2-29 was missing from my copy so I do not know what this memo was). Thus the numbers were taken in the range 0.01 to 0.3, conditional on no depressurization, which, itselfis conditional on core melt, was taken as 0.2. First, a formal matter. I agree with BNL that if the experts were asked to quantify these matters for Seabrook they would, in all likelihood, come up with the same numbers. Still, at this point the BNL report should explicitly state whether the BNL team examined the Surry-Seabrook relationship with regard to the phenomena ofinterest here and what were their conclusions.
Now, on the substance of those numbers. It is not stated whether the 80% chance to depressurize includes operator actions for doing so or whether the 0.8/0.2 split simply reflects
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the experts judgement on the likelihood to fail the primary system prior to core melt, and elsewhere than the steam generator tubes. If the former was a significant consideration,I l
would call it into question on the basis that current systems and procedures cannot provide any l assurance one way or another (it they did we would have no problem). If the latter was the l significant consideration the numbers represent a pure guess which I would also call into i
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question. In my view the way things presently stand system failure is a vinual certainty. The 4
real question then is what fraction of those failures could be in the steam generator tubes. I '
believe one would be hard pressed at this time to show a chance less than of order 1. In
~ fact, considering that certainly not all tubes would be in " mint" condition, and lack of detailed data and structural analysis under the relevant conditions, I would not be very surprised if this number turned out to be even larger. Thus I do not feel comfonable with the comparable BNL range of 2x10-3 to 6x10-2. In my view, for this kind of study, use of numbers less than 10-1 requires substantial understanding of all technical elements. I think pan of the problem is that -
the process has been broken, anifically,into two parts, i.e., the numbers would have been even smallerif a three-pan process was considered.
In summary, then, a prudent estimate of SGTR probability would be of order 1 Combining this with (the best estimate I suppose) frequency value quote 9 as yields4x10-5an estimate of 4x10-6 for bypassing containment (assuming that secondary side behavior was included in the 4x10-5 value). Assuming one order of magnitude uncertainty (I would not be i surprised ifit was more) in the above frequency estimate and a factor of x2 in the conditional rupture estimate we obtain an "uoper" estimate of 4 ner reactor year. I think this is an unaccentably larce number.
Section 23 (pp. 2-27 and 2-28) provides, in brief, the source term methodology employed and resulting consequences. Release category SIW (large early bypass) was chosen as a conservative assumption. I agree that this is conservative, perhaps too conservative.
Some retention during blowdown would undoubtedly occur; however, it would strongly depend on the rates of blowdown, past history in the sequence, and details of the kinetics of l
deposion and resuspension processes. None of these is well known, and I can sympathize with the BNL team for not making arbitrary cunai!ings in this source term dermition.
The discussion of consequences provides a general perspective that SGTR does not dramatically alter the risk even if what the report calls pessimistic assumptions were to be used.
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This is a result that needs a lot of explaining. It implies, basically, that early-containment failure at the assumed scenario frequency, is not such a big deal and I certainly disagree with this perspective. Furthermore, here is the place to discuss uncertainties in frequency estimates 1
and their impact. Funher, the statement on what the NRC staff " believes" is nebulous and only serves to incorrectly diffuse the issue.
To summarize. mv own ficurine (as explained above) indicates that allowine hich oressure scenarios to nroceed unchecked is unaccentable. This judgement is based on NRC safety goals, on the perspective provided by the risk estimates in the BNL report together with my
" upper" estimate of containment bypass (~ 104 per reactor year), and consideration of j intangible aspects regarding public response to such potential accidents.
' 1 Section 6.4 summarized the overall sensitivity study results. There are several difficulties here. -
(a) The " combination" of sensitivity study results is strongly qualified that is "not rigorous and could lead to inconsistencies." This is like disavowing the results altogether. If something is salvageable the authors should explain what it is and why it is useful. Explaining clearly how i
results were " combined" also would be necessary-I suspect they were simply added up.
(b) The results are given in graphical and tabular form, cut-and dry, with no explanations.
Furthermore,I see no attempt at interpretations. One may argue that this is the job of the NRC' ,
but, I believe, the analysts are best qualified in judging, at least, whether differences are sig'n ificant. Furthermore, major trend differences should be discussed and explained.
(c) 'Ihe final paragraph asserts that "the conservative assumptions regarding accidents during shutde.un and induced SGTR have the most impact on the dose vs distance and risk estimates in PLG-0465." This does not say whether in the BNL team's opinion this impact is significant. I do not know whether this noncommittal position was intentional or simply a poor choice of words. There is a good chance it may be the latter, especially since it goes on to say that the optimistic assumptions have a min.cr impact and that there is considerable uncertainty in l
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the areas explored in the report (implying that the " pessimistic" estimates entail a significant impact). -
(d) If my deduction in (c) above is correct then there is a need to explain how the combined effect of a few rather menial (individually) effects can be so significant. That is, when uncertainties cannot be discussed properly even on an order of magnitude basis are factors of x2 in the individual results significant? and is the non-rigorously formulated " combined effect" significant? Of course, I do not believe the individual area of SGTR is insignificant.
Furthermore, I do not think the subieet of DCH was cronerly treated nor do I think it is not sienificant.
(e) If my deductions in (c) above are incorrect, the BNL team should indicate, more clearly, that the final result is insensitive to the assumptions in the sensitivity studies. Again, I would disagree with the conclusion, however.
- 3. Comments on the NUREG report My key comments on this report will be categorized under the headings: Perspective, Philosophy, Mechanistic Assessment, and Implementation. Additional comments will be given under the heading " Details."
l 3.1 Question of Perspective
' There is a repeated reference attributing the genesis of the SGTR topic to a " strong" I
, Seabrook containment and hence the unfolding of"previously neglected bypass paths that were masked and found now to contribute to risk" (p. 5, p.11, etc.') I think this is an unnecessary
- finesse of the issue at best, but badly inaccurate at worst. The simple fact is that the potential l
l for this particular mechanism to fail containment was discovered only within the past two years and it is not fully appreciated to this day. Furthermore, the historical perspective provided in i
i the report is rather incomplete. I suggest that instead of dancing around the issue (a lot of
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8 unnecessary pages in the report on that ) it should be clearly and blantantiv introduced. tocether with the key milestones and references in its deveicoment. alone the lines shown in my introductory section.
3.2 Question of Philosophy The report gives an unmistakable impression (i.e., p.11) that the Regulatory process
. operates in a never-ending fashion. That is, as plants get stronger we get more picky. This is a wrong and highly undesirable impression. ,Rather, it must convey the impression that assessments are made considering all (that we know) competing risk factors in as absolute sense as it is possible The report acknowledges that it did not make an effort to assess the pros and cons regarding the need to depressurize the primary system in case of such accidents. This is a short-sighted view and I strongly recommend that it be revised. In my view the benefits of depressurization are straightforward, the costs are minimal, and there are no downsides to it.
Along the same lines, the whole question of operator action is 1:ft hanging. In the absence of explicit procedures we should not count on an operator thinkine that it would be good to depressurize the system. And how will he obtain valve controlin a station blackout?
3.3 Questions ofMechanistic Assessment -
The key point made in the report is that experimental evidence and its applicability are preliminary, that the applicant's Computer Program has not been verified nor documented, an that certain modelling assumpdons are optimistic. I agree with all three. In addidon to these purely thermal-hydraulic aspects there are other more serious quesdons that are not even j touched. These include: (a) a whole range of issues regarding structural response under l
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- highly complicated, very far from design condidons, (b) uncertaindes regarding actual steam generator tube conditions (certainly they will not be " mint") at the time of the postulated 3-r - r--- , e-sa- g- -.,--- w-- y ~y - wy- , - ---v.,-----m- - - - - . --g,----,---w----g- - - - , ---w,, wuw.--uww,, v-,w-y ---
9 accident, and (c) major uncertainties in fission product release and deposition (or a whole sequence of such phenomena) processes.
Concerning the assumed plant state the report correctly identifies the applicant's assumption, that the secondary side will be at 1,100 psi, as inadequate. After all is it not the case of bypass (open secondary side to the atmosphere) the case of real concern? Along these lines the report should also discuss what can happen if the tubes fail with a pressurized secondary side. For how long will the relief valves (on the secondary side) lift? How much steam will be expelled? What is the likelihood that they will fail? What is the likelihood, in case of a massive SGTR, that the secondary system will fail somewhere? ~
All calculations have, presumably, assumed intact loop seals. In my opinion SGTR t
becomes a certainty in the absence ofloop seals. Hence, a comprehensive, fully-documented, assessment ofloop seal behavior under the whole range of possible scenarios is essential here.
3.4 Questions ofImplementation The report gives the optimistic impression that this problem area can be " studied away." I believe this is an overiv ontimistic and inanprontiate iudgement for the Regulatory to make at this time.
Furthermore, the procedure outlined to " substantiate this judgement" is unrealistic and incomplete. How are we going to guess the structural condition of all steam generator tubes at the time of a hypothetical accident? How are we to verify structural evaluations under such highly overheated conditions? What experiments will give us adequate data to ensure proper verification of the fission product migration process? To be sure, this last comment refers to additional heat loads on the tubes rather than to source term effects. 'Ihat is not to mention verification of scaling for any natural circulation experiments and associated numerical tools.
How long and how much money will it take to obtain a sufficiently good " substantiation" of thisjudgement?
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10 3.5 Some Additional Details. ~
- 1. p.15. He whole discussion at the top of this page is misplacing the issue. Only RCS pressure matters and item 3 is particularly irrelevant.
- 2. p.16. The major contributor here is station blackout and there is no power to drive the pumps. He comment at the bottom, p.16 and top, p.17,is thus irrelevant.
- 3. p. 20. Reference is made to section 2.2.4 but there is no such section. Here it should state on what basis can PSNH count on operator access to PORVS and for what cases they cannot.
- 4. p. 21. The discussion here on loop seals and pump use is appropriate. .
- 5. p. 22. Here the authors of the report visualize removal of a loop seal"as water was forced out of the RCS via the break." his is not possible for such low rates of steam f' low.
However, as superheated steam kept bubbling through the loop seal water, it would slowly boil it off, having the same effect. Here one must be concerned with a race between system depressurization, loop seal evaporation, and steam generator tube heat up. After the loop seals break the steam generator tubes would nearly equilize with core temperatures very rapidly. -
- i 6. p. 34. How about failure modes other than creep ruptures?
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- 7. p. 36. Here is need to go much more into the PSNH estimate of 10-2 to 10-3 per demand to fail to depressurize the system. Here is where the major pay offis and here it is that
'some considerable progress needs to be made.
- 4. Conclusions The NUREG report is significantly limited in its perspective and philosophy in approaching this problem. De BNL study on the impact of SGTR is highly qualified and in my opinion does not bring out the problem as forcibly as it should. It is promising that the utility believes that they can reliably depressurize the system. I believe this is the most I
11 pronusing avenue and it should be pursued aggressively. The benefits will be farreaching not l only for Seabrook but also for all other plants. Instead of this limited report I would favor a comprehensive one dealing with all aspects of high ressure P scenarios and an assessment of needs and procedures for depressurization.
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DEPARTAIENT OF CHEh!! CAL AND '
Ntl CLEAR ENCINEERINC 5ANTA BARBARA. CALIFORNIA 93106 1
November 21,1986 Dr. T. Speis, Director Safety Technology Office of Nuclear Reactor Regulation US NuclearRegulatory Commission MailStop P-1122 -
Washington,DC 20555 -
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Dear Dr. Speis:
! Re: Direct Containment Heating, etc.
Nearly three years ago,in the NRC/IDCOR meeting dealing with the Direct Containment
' Heating (DCH) problem I presented some simple calculations that led me to make the follow
-deductions and recommendations.
-1.
High pressure steam is very effective in redistributing the core heat throughout the primary system. That is, different parts of primary systems follow the core temperature within a few hundred degrees centigrade, which implies failure of the primary boundary prior to gross core meltdown. As a consequence, all high pressure scenarios should be expected to revert to low pressure ones.
- 2. Prediction oflocation of failure is rather difficult and highly uncertain. 'Ihis is particularly so for the steam generator tubes, which may be in unknown states of degradation. Such failures could cause containment bypass (i.e., in-station blackout situation) at a rather i inopportune time regarding outside consequences.
- 3. ~ Given the above it would appear prudent to ensure timely, reliable, and predictable primary system depressurization. This can be accomplished through a variety of means, including
" fuse" systems or automatic (or manual) depressurization systems with dedicated power supply (or steam driven). The selection should be left to the utilities.(vendors).
- 4. Given item 2 above, the burden of proof for requiring such a system is not with the NRC.
Rather, the burden ofproof for not having it is with the licensee.
Subsequent calculations of our own as u ell as such done by EPRI, INEL, LANL and SNL seem to support the view that early failure (prior to gross core melt)is very likely. Also, experiments run for EPRI at Westinghouse further demonstrate that steam generator tubes would be involved in this heat up process even in cases where the loop seal had not cleared.
Since my original suggestion I spoke often about my views on this subject. I also expounded on them during a seminar to the ACRS of" Severe Accident Phenomenology for PRA use."
Now I have the following additional points to make.
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November 21,1986 Page 2
- 1. '
Meanwhile, the DCH issue has persisted-in fact it has escalated. High pressure dispersal has been pursued at SNL for over 4 years now and new work has begun at BNL, with a significant commitment of resources. I have not heard yet from either of these two research teams an engineering judgement that dispersal yielding a DCH contamment failure is a figment of the imagination. On the contrary, all signals I can read in meetings, etc., indicate a real concern in this area.
- 2. De installation of a system assuring reliable and predictable depressurization would instantaneously remove all DCH concerns and associated expenditures. By nature the DCH problem is like that of energetic steam explosions, i.e., hard to obtain the kind of evidence needed to satisfy those more skeptically inclined. His of an engineered depressurization behavior. even However,provides an additionalincentive in f if the DCH problem, somehow, went away, the potential for steam generator tube failure would still require that the action recommended here be taken anyway. -
- 3. It is my understanding that others (notably Jesse Ebersole of the ACRS) have ar favor of a depressurization system in PWRs for reasons other than those stated Such here.gued in potential additional benefits should be factored into the decision makmg process.
He purpose of this letteris to express and document my opinion that this problem area needs .
to be addressed by the regulatory authorities with the utmost urgency. Please feel free to call on me ifI can be of any help.
I j Dank you foryour consideration.
Sincerely, i
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T T.G. neofanous, Professor l
! Depanment of Chemical and Nuclear Engineering Director, Center for Advanced Multiphase Processes and Safety TGT/h cc: W. Morris C. Allen F. El Tawila
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NUCLEAR ENCINEERING SANTA BARB ARA, CALIFORNIA 9JIO6 December 8,1986
- Dr. T. Speis, Director
. Safety Technology Office of Nuclear Reactor Regulation US NuclearRegulatory Commission Mail Stop P-1122 -
Washington,DC 20555
Dear Dr. Speis:
Re: High Pressure Scenario, etc.
1
- 'Ihis is a followup to my November 21,1986 letter conceming my recommendation t
j~ a regulatory requirement for the timely and reliable depressurization of all Light Water Reactors
, under all possible severe accident conditions including station blackout. In my previous letterI i;; expressed expand on this this asconcern from the Direct Containment Heating standpoint. Here I would like to recommendation.pect and to introduce some additional considerations that further support my
. 1.
Current approaches to the DCH-dispersal problem are based implicitly or explicitly on assumption that venel blowdown will occur from a single instrument tube penetration failure.
i I believe coherent failure of many such penetrations is not only possible but rather likely the assumptions of high pressure meltdown phenomena. Currently there are no reliable methods to estimate the coherence of such failures, and considering the superposition of uncertainties in the phenomena leading up to this stage it is highly unlikely that any but gros bounding estimates would be possible in the foreseeable future. On the other hand the effects on dispersal can be quite dramatic. We have carried rough calculations of dispersal potentia the Zion geometry and concluded that this item alone can dictate behavior. In addition to impacting strongly the velocities in the Steam Generator compartments (where de-entrainme should occur if dispersal is to be avoided) there is potentially an important effect in sca phenomena within the reactor cavity (i.e., dimensions of entr'ained particulate). We are currently addressing quantitatively both issues; however, I bring them up here because they associated with important and inherent uncertainties and thus have to be weighted in the overal approach of addressing the DCHissue.
2.
Even if dispersal was eventually shown to be a non-problem the high pressure blowdown would be undesirable from a hydrogen behavior standpomt. To fully achieve the benefits of igniters, both in terms of their performance, as well as in terms of our expectation of such benefits, we must avoid,I believe, rapid releases oflarge quantities of hydrogen--as would potentially occur in a high pressure scenario blowdown. This is particularly so for ice condensors, but large dry containments could be also affected.
- Dr. T. Speis December 8,1986 Page 2
- 3. Even the steam explosions area would be favorably impacted by avoiding the high pressure scenario. Dere are two aspects here. The one arises from the loss of strength of the upper internal structures and vessel head bolts under the steam natural circulation that accompanies the high pressure scenario. Over 500 MJ of mechanical energy credit would thus be lost in the energy partition evaluations of potential missile energies. Dat is, much smaller steam explosions (even those that do not fail the lower head) could potentially create missiles capable of o failure, ne other aspect arises because of the increased quantities of premixing expected in high pressures and confirmed by ourrecent calculations. Although there are some doubts concerning the escalation of triggers to explosions at high pressure it would not be pmdent,in my opinion, at this time to discount the likelihood of such explosions, ne structural aspects
- of the above have been documented in our four recent steam explosion papers. He premixing aspects at high pressure are presently written up in a separate paper.
- 4. Finally, let me not forget that timely depressurization might even enhance the recovery -
potential. This is because it would bring into action the accummulators with good potential to quench the core and thus buy additional time for recovery of power.
In Conclusion Here we have a perfect example of how a major portion of the perceived risk as well as of the resource commitment necessary for its full appreciation (not necessarily alleviation) can be eliminated through a rather straightforward system change. I believe the matter needs the most urgent attention by the Research and Regulatory authorities. From my discussions with
-l " Systems people" I am convinced that implementation of a highly reliable, manually activated, j system would be rather trivial. A variety of fuse-type devices (passive) are also possible. I
- think the NRC should request ideas from the utilities and vendors and select the best one
, i among them. One could even make a design competition-with awards--to generate additional incentives. Please let me know if you have any questions or ifI can be of any help.
ll Sincerely, ll 1: - _
T.G. Theofanous, Professor 7 i Department of Chemical and Nuclear Engineering Director, Center for Advanced Multiphase Processes and Safety TGT/h cc: E.Beckjord D. Ross W. Morris C. Allen F. ElTawila D. Powers
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DEPARTSIENT OF CHEhllCAL AND SANTA BARBARA, CALIFORNIA 93106 NUCLEAR ENCINEEDINC December 11,1986 Dr. T. Speis, Director
- Safety Technology Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission MailStop P-1122 Washington,DC 20555 -
Dear Dr. Speis:
Re: Depressurization System for PWRs Following up my December 8,1985 letter I would like to clarify that there is considerable flexibility in specifying the timing and duration of the depressurization operation. This flexibility, of course, is extremely important in meeting the overall objectives with a set of design criteria for the depressurization system that are both easy and inexpensive to meet.
Funhermore, because of this flexibility it should be easy to achieve the high reliability necessary while minimizing the consequences ofinadvertent actuation.
Sincerely, T.G. Theofanous, Professor g Department of Chemical and Ndelear Engineering Director, Center for Advanced Multiphase Processes and Safety TGT/h I
cc: E. Beckjord .
D. Ross B. Morris C. Allen F. El Tawila D. Powers l
Machmed 'l
- UNIVERSITY OF CALIFORNIA. SANTA BARBARA ,- .*
BtRn;ELEY
- D tvis
- IRYl%E
- LOS ANCELES
- RIVER 58!)E
- SAN DIECO . 3 4N FR ANCISCO
$ANTA BARBARA
- SANTA C3tt'Z y , -
DEPART 3 TENT OF CHEntlCAL AND S ANTA BARBARA, CALIFORNIA 93306 Nt' CLEAR ENCINEERINC 1
December 19,1986 Dr. W. Morris Office of Research U.S. Nuclear Regulatory Commission Washington,DC 20555
Dear Dr. Morris:
Re: Further Thoughts on Depressurization.
~
l We have carried out some simple calculations to illustrate the flexibility in aming and duration l
l of depressurization needs mentioned in my December 11 letter. The primary system volume 3
- was taken as 270 m . Steam pressure and temperature were taken at 15.5 MPa and 445' C (100* superheated), respectively. The system pressure transients for a spectrum of vent areas are shown in the attached figure. Note that for typical designs the accumulators will come on at
- 4 MPa condensing a great deal of the remaining steam. Thus, a vent area of- 6 in2 should prove more than enough. Clearly, there should be no problem providing controlled venting of this size with the required high reliability.
Even if it took 15 minutes to depressurize, the core would sustain no significant clad oxidation damage (adequacy of vent area, as small as 2 in2 ,in this respect would also appear conceivable). Thus, full core quench, with attendant repressurization would be expected. A new boil off, heatup cycle would then follow, for which the depressurization characteristics mentioned above should also be quite adequate. This time extension would provide additional opportunities for recovery of power and termination of the accident.
I recommend that more detailed depresstMzation transient calculations and associated scenario determinations be made as part of the design effort (by vendors and utilities). I also recommend that you appoint a committee consisting of D. Powers, J. Kelly and myself to cany out a similar assessment for the NRC.
i Sincerely,
( -
T.G. Theofanous, Professor Department of Chemical and Nuclear Engineering Director, Center for Advanced Multiphase Processes and Safety TGT/h Attachment cc: E.Beckjord C. Allen D. Ross F. El Tawila T. Speis D. Powers i
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)hhment 6 Sandia National Laboratories Albuquerque. New Mexico 87185
. December 24, 1986 Professor T. G. Theophanous ,
Dept. of Chemical and Nuclear Engineering University of California, Santa Barbara Santa Barbara, California 93106 i
g-
Dear Theo,
i Thank you for sending me copies of your letters to Dr. Speis of November discuss PWR 21, depressurization.
December 8, and December 11, 1986, in which you I wish to express my agreement with your view that depressurization of pressurized water reactors under accident conditions could be desirable.
~You note, accurately, that there is an escalating concern about ~
' sion directofcontainment core melt. heating as a result of pressure-driven expul-Though it is true many have calculated that natural convection within the primary circuit might cause system depressurization, the experimental evidence of the Three Mile Island experience makes it difficult--actually impossible--to conclude pressure-driven melt expulsion from the primary circuit vill not occur. I do not believe calculational efforts have yet given sufficient credence to phenomena that may disrupt the-distribution of heat to the primary pressure boundary. But, even 1 if these -calculations are entirely accurate and the TMI-experience can be shown to be a fluke, there is the problem you note of the inherent unpredictability of the rupture point.
Exchanging direct containment heating accidents for v sequence accidents is a poor trade-off.
Clearly, it would be preferable to avoid, altogether, the pressurized accident sequence. Certainly, with the exception of i
some station blackout sequences, BWRs have benefitted from a severe accident perspective because of assured depressurization.
.You PWRs.
note I in your letters several benefits for depressurization in perspective. believe there are other benefits from a source term L
Current modeling in accident analysis ~ codes of fission product release is quite crude.
i as imp ~ortant variables. It only considers time and temperature As a result, the~re have been difficulties predicting results of tests. In connection with both the MELCOR and the MELPROG codes, we at Sandia have been upgrading the fission product release modeling. One of the upgrades has been to include rate limitations caused by gas phase mass transport. When applied to pressurized accident sequences, gas phase mass transport processes are found to sharply limit the release of the more volatile fission products--Cs, I, and Te.
This is not good. If these volatile fission products are not i
i
. . - . - - . - ~ _ . . - - - - . . . . . . . - _ . - . - . . . . _ - - . - . - -
- [-l***ProfessorT. G. Theophanous December 24, 1986 released in-vessel, they will be released ex-vessel, and ex-vessel release is not attenuated by deposition in the primary piping system.
The situation is even worse if the ex-vessel release is the result of direct containment heating processes that also fail the containment.
We have also been upgrading the models of the chemistry of radionuclides during core degradation. In this work, we have found high pressure steam will augment the volatility of many of the less volatile radionuclides:
[Mo]UO2 + 4H 2O H2moo 4(gas) + 3H2
[B40]no 2 +HO2 Ba(OH)2(gas)
UO2 + 2H2O H2 UO4(gas) +H2 The steam augmentation of the release of these radionuclides w'ill place additional heat loads on the primary piping system. In particular, thermal loads on steam generator tubes may be enhanced--again threatening sequence V type events. We suspect, but have not proved that high pressure hydrogen will also enhance the volatility of radionuclides now thought not to be released during the in-vessel phases of an accident.
Assuredly, the complexities of both the chemistry and the release of radionuclides caused by high pressures would be obviated if the PWR were depressurized by some design system. The existence of a known flow path from the primary system would permit useful engineered mitigation of in-vessel fission product release. Were the vent into a containment sump, then release would be attenuated by the same mechanisms that attenuate release during blowdown through a BWR suppression pool. May I add, then, to your suggestion of PWR venting that this ventino be done under
[ water because of source term considerations.
I tioncannot comment on the ease of introducing assured depressuriza-of PWRs. It does appear, however, that the idea was merit and does deserve some attention.
Sinc ~erely yours, h oW Dana Powers, Supervisor Severe Accident Source Terms Division 6422 Copy to:
USNRC - T. Speis USNRC - D. Ross USNRC - F. Eltavila
"- .. jeachmed 7 EPRI NP 4455 React raccidents 1 8 d Pro eedings Electric Power fgurce e, core accidents Research Institute March 1986
_~
Proceedings: The Sixth
- . Information Exchange Meeting K. $ y &4 on Debris Coolability mWMW 20 -
9[ cdocLRtb NO V. 7 Cj , (g gg .
L Prepared by University of California at Los Angeles Los Angeles, California
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l Section 24 NATURAL CIRCULATION PHENOMENA l' AND PRIMARY SYSTEM FAILURE IN STATION BLACKOUT ACCIDENTS
. by H. P. Nourbakhsh, Chien-Hsiung Lee and T. G. Theofanous School of Nuclear Engineering
' Purdue University West Lafayette, IN 47907 9
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INTRODUCTION
',3 E The potency of high pressure steam natural circulation phenomena ue to re i
decay heat within the primary system components during the re- core post-melt period has not been fully appreciated in the past.
The original suggestions
- i of their possible significance were made by Denny and Sehgal [1] an Denny and Sehgal built an idealized natural circulation flow regime
, coupling the I- core with the upper internals, into the code CORMLT.
From calculations for a Station Blackout accident that are emphasized to be preliminary, they d cant implications concerning: -
delays in core degradation, reductions in thermal energy released to containment, and higher hydrogen generation rates. Winters extended Station Blackout calculations with RELAPS into the post-dryout regim i
These calculations produced a natural circulation loop between the reactor and the steam generators and an associated energy redistribution of such ma ,
that the possibility of primary system boundary failure prior to fuel cladding
. failure became evident.
2 Theofancus and Lee [3] also considered this mechanism in an ass likelihood of the so-called "high pressure scenario," i.e., vessel failure and core release to containment from a high primary system pressure. This simplified analysis will be sum.arized first as a way of developing a feel for the order of magnitude of various effects.
Some simple experiments of natural circulation within a partially volumetrically heated porous medium will be presented to more concretely demonstrate.the efficacy of natural circulation cells to penetrate df.oply into a heated porous bed.
This paper will then conclude with a brief de-scription of a numerical model including applications to this demonstration ex ment as well as the reactor conditions of interest.
i d
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24-1
7
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ORDER OF MAGNITUDE ANALYSIS We consider a simple 3-mass lumped parameter model as illustrated in Fhure 1.
Each mass is characterized by the same temperature and the fluid leaving is sup- -
posed to be in thermal equilibrium with it. This is a reasonable assumption for this approximate analysis given the highly distributed (high interfacial areas) core, upper internals, and steam generator masses. The coolant flow through each mass em , is controlled by a loss coefficient (permeability) and the available head due to the temperature and associated density differences.
. w---- 3 -- w 2
h 1as __ 2 s In o 18 ,
I1 3 --- :-
Figure 1. Schematic illustration of the 3-mass lumped parameter model.
The mathematical model may be expressed as:
i di R
dt
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=
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3 (4)
_A p
where R. = , and Q' =
(5) c'pc c pc 24-2
, ,. ., j j .. -.
o' The solution for this system is: .
~
(6)
. T, = C e*1# cosz,t + C2 8 1' sinzz t + Cs t + C.
3
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(7)
+ C, + R,(C,z, - C,z,) e*2# sinz,t + C,t + R C, + C, 3 C, R, R, T, =O't7+ -
T, - T, (8)
R1 R R 1 1 1
where
. C, = T,(R,+R,+R,) ,
C, = T, - C, . . C, = - C,z,/z, - C,/z, 2
C, = Q'/cR , C, = C,/c R, - bQ'/c R, .
a = R,R, , b = R, + R, + R,R,/R, , c = R,/ R, + R,/R, + 1 2
z, = ~ b/2a z,= (4ac - b )1/2/2a For the computations Eq. (4) was coupled to Eq. (6) through the fluid equation of state and solved by successive approximations. Several numerical examples were considered to determine the effect of the assumed natural circulation flow pattern and of the core pemeability (which would be related to the degree of degradation assumed). Geometric and mass data used are typical of a Westinghouse 4-loop plant.
The In all cases primary steam was at 2300 psia and decay pcwer was set at 42 MW.
results, are summarized in Njures 2 and 3.
For the " loop circulation" case it is assumed that the loop seals at the pump and lower plenum locations are broken and a steam flow path is open all around the primary system. The masses m , m,, and m 3 3
, are identified with the core, upper internals including upper head, and steam generator masses respectively. No heat los'ses to the outside of this three mass system are cor.sidered. Frictional losses in the core region dominate and they were represented by a permeability value of 1.5 x 10~' m 2 which corresponds to a rubbled core with a characteristic particle dimension of 8 m and includes a factor of 1/2x for deviations from spherical shape.
A parametric using a higher permeability by a factor of Sx was also considered to The calculations yield a nearly steady represent a less disrupted core geometry.
natural circulation flow of 25 kg/s per loop. This flow is sufficient to redis-tribute the decay heat such that the three mass system heats up after an initial transient of -1 hr., essentially with a uniform rate of -1.7 x 10 'C/s (Figure 2).
The upper internals follow closely the core temperature while the steam generator 24-3
- - - .,. m __
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lags by -150, 70'C for the two cases considered.
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- i. - -
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LJ Figure 2. Predicted temperature tran-sients for the " loop circulation" case. (q[ 8
- % ; ; . io t x t o'* (.)
Figure 3. Predicted temperature tran-sients for the "in-vessel recirculation" Case.
p ..
Although some code calculations [2] do yield opening of the loop seals it is generally recognized that significant uncertainties in predicted behavior exist in this area. To complement the behavior depicted in the previous paragraph in this regard the "in-vessel recirculation" case was also considered. The flow pattern envisioned is shown as an insert to Figure 3. Now the three masses of our model
~
were identified as the core, one-half of the upper internals (central region), and the other one-half of the upper internals (outer region). The same two cases of core permeability as in the. loop circulation case were considered. For the nominal case a gradual increase in natural circulation flow to a steady value of 45 kg/s over a period of 2000 s was calculated. Again, these flows provide substantial thermal coupling between the core and upper internals masses. However, as seen in Figure 3, this coupling is not as efficient as in the loop circulation case. That is the rate of heatup is different for the three masses and at 10,000 s the outer portion of the upper internals lags by 120 and 60*C behind the core temperature for the two cases considered. Furthermore, as illustrated in Figure 3 these high 24-4
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temperatures would propagate into the hot leg, towards the steam generator, and
{ pressurizer lines, by net flow caused by relief valve cycling as well as_by
.-h counter-current steam flow in the hot leg [4].
These scoping studies indicate that primary system temperatures would follow the core temperatures during heatup within "a few hundred degrees Celsius" such that primary system boundary failure prior to core melting and slumping would be expected. Such failure would be due to loss of strength of structural material expectec at around 650'C, i.e., would produce leak areas sufficient to depressurize the system. Furthemore if conditions for the loop circulation case were to be present, concerns about steam generator tube failures (by the same mechanisms) at high pressure should be addressed. On the other hand, conditions leading to clad ballooning to such an extent that blocking of the flow paths in the core occurs
. could also significantly alter these predictions. In such event phenomena of clad oxidation, melting, and relocation (which will reopen the paths), and fission product relocation within the primary system (and of associated themal loading) '
must also be considered together with the natural circulation flows discussed here.
AN EXPERIMEfiTAL ILLUSTRATI0ft Purdue's Large Scale Debris Bed Coolability Simulation Facility [5,6] was adopted to illustrate the energy redistribution mechanisms for the "in-vessel recircula-tion" case. The present bed measures 21 cm in diameter and 100 cm in height and is packed by alternating layers of 1.25 cm aluminum spheres and a layer of similar thickness of 0.8 cm irregularly shaped stone fragments (gravel). It has a porosity of 3.86 and a turbulence permeability of 1.4 x 10" m 2 which is close to the permeability of a bed made up only with gravel and about one-half that of a bed of equal size spheres and the same porosity. The power to each layer of aluminum spheres is individually controlled such that any vertical portion of the bed can be volumetrically heated at will. For the experiment described herein a total of 5.7 kw were applied over a bed height of 30 cm extending from an elevation of 23 cm (from the bottom of the bed) to the 53 cm elevation. The bed was flooded with 20'C water and the experiment lasted until the bed reached boiling. The tempera-ture transients were measured by means of 250 thermocouples appropriately distri-buted throughout the bed. These themocouples were scanned every 5 s with the help of a PDP-11 minicomputer.
The results are shown in Figures 4 and 5. From Figure 4 we can clearly observe the conduction regime and a rapid transition to convection which is of sufficient strength to quickly homogenize (themally) the whole bed. The prediction of the simple model already discussed for which convection was taken to initiate at a critical Rayleigh number of 30 is also shown for comparison. Observed spatial 24-5
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temperature distributions at selected times are shown in Figure 5. This figure contains also predictions based on the numerical model described in the next section, too , , , , ,
T
- re -
g T,
, 3 .. . . .
H se . . . .
,e a e e e io is -
t a t o-a ,
Figure 4. Comparison of 3-mass lumped parameter model predictions with experi-
! mental results.
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2 .
s ... .
z .
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- e 0.3 e 0.8 0.7 ) s.0 Z/H Figure 5. Comparison of numerical model predictions with experimental results.
24-6 .
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. A NUMERICAL MODEL The "in-vessel recirculation" case was further examined by means of a di.stributed parameter model as follows. The whole vessel is viewed as occupied by a porcus .
medium, the region corresponding to the core being heated while that corresponding to the upper internals being inert. The pore space is occupied by steam which is The vessel, as a vhole free to convect as buoyancy forces develop from heating.
is adiabatic. friction is governed by Darcy's law, and density variations are handled in the Bussinesque approximation. This model with the e.xception that it includes themal nor. equilibrium between the fluid and the particles is commonly used for natural convection in porous media and may be expressed as:
" (9)
+ + =0 ar .
(10) h- + p u = 0 .
(11)
+ 04 9 - S (Tg - gT ) g go + "* =0 BT (1 - c) o c a[- = h - h S(T p -T) g 2
2 3T 8T 1 BT P + + (l2)
+ kp (1 - c) ar BT ST
=
cog e g 3[ + up ca OT)*VPc fg 2 2 BTg)
(13)
= hS(T P
-T)+Kcf3Tb+3TE-+1 5 6 L 3r8 By:
r 3r J This model The boundary conditions are obtained for impermeable adiabatic walls.
l was solved in nondimensional form by the finite difference method using upwind differencing and successive over-relaxation. At each time step the energy equations were solved explicitly (with implicit coupling of the* exchange terms), while momentum and continuity were solved simultaneously using a marker and cell technique.
For all calculations reported here a number of nodalization, time step, and con-vergence criteria, studies were performed to assure the reliability of computations.
Clearly, appropriate porosity, permeability, and power distributions may be used in detai' led evaluation. Two cases are considered here.
I 24-7
-.=~.~-m;rn m m s,27-,-[M M p _
.- s' *o f
Calculations for the demonstration experiment discussed in the previous section were carried out using the measured bed porosity of 0.386 and a permeability of 1.5 x 10-' m 2. From the comparison of' Figure 5 we deduce that all essential .
features, including the conduction heatup regime, the time for the onset of convec-tion, and the rap 1 transition to uniform heating following the onset of convection, are adequately predicted. In addition, for a short period following the onset of convection the model predicts a turnover similar to that observed in thermals with sufficient cold fluid penetrating the heated region to temporarily invert the ver-tical temperature gradient. Such behavior was qualitatively observed also in the experiments. However, due to non-axisymetric behavior in the experiments precise interpretation of this aspect must await further investigations.
Reactor calculations were carried out with a uniform porosity of 0.5, a core power of 42 MW and a radial power distribution given by:
= 2.32 Jo (4.81 h ) (14) ,
2 A low permeability value of 1.5 x 10-' m was chosen to maximize the thermal gradi-ents calculated. An intact core geometry would be characterized by significantly '
higher permeabilities. Selected results of one such calculation are shown in Figures 6 and 7. In figure 6 we observe the unicellular flow pattern envisioned in our simple model presented earlier. In fact the predicted flow rate is -50 kg/s which is in good agreement with the recirculation rate of 45 kg/s produced by the 3-mass model. The time-wis.e variation of the spatial temperature distribution c,ay be visualized as in Figure 7. The isothems are nearly vertical and seem to be continuously (in time) displaced outwards as new higher temperature isotherms are generated near the centerline. However, the maximum temperature difference at 10,000 s is only 240*C, and most of the mass is within 1%*C which is in good agreement.with the 120*C obtained in the 3-mass model.
i CONCLUSIONS l
High pressure steam natural circulation phenomena can be responsible for redistri-buting the core decay power to all primary systems components that are accessible to form circulating loops with the core region. As a result the primary system boundary is expected to fail prior to core melting, and the so-called "high pressure scenario" would appear unlikely. T5e possible role of clad ballooning, and oxidation remains to be assessed before these conclusions can be applied to l all possible variations of a Station Blackout (or other similar) accident.
24-8 I
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Figure 7. Isoterms at 1416, 2832 and 9912 s.
ACKNOWLEDGMENTS The financial support by the U.S. Nuclear Regulatory Commission under Contract No.
NRC-03 83-093 is gratefully acknowledged.
i 24 9
~
l q
^*
_ s s .. p .
a4 NOMENCLATURE A
= flow area
.C .Cg = specific heats of coolant and of structure '
D = vessel diameter -
g = acceleration of gravity h = heat transfer coefficient between fluid and particles k ,k fP = thermal conductivity of fluid and effective thermal conductivity of porous medium m = coolant flow rate m = mass of i U compartment p = pressure Q = volumetric heat generation rate r = radial coordinate .
S = the surface area between the fluid and porous medium per unit volume .
Tg = temperature of 1 0 compartment t- = time u = velocity component in r-direction v = veloi:ity component in y-direction y = coordinate in vertical direction Greek Symbols 8 = thermal expansion c'oefficient p = average density of fluid
, u = viscosity -
n = frictional length / elevation length 6 = h,3/h33, where h 33 and h 33 are shown in Figure 1 x = permeability of the porous medium REFERENCES
- 1.
- V.E. Denny and B.R. Sehgal, " Analytical Prediction of Core Heatup/ Liquefaction /
Slumping." Paper TS-5.4, Proceedings Intl. Meeting on LWR Severe Accident Evaluation, Cambridge, MA, Aug. 28-Sept. 1 (1983).
- 2. L. Winters, "RELAp5 Station Balck-out Transient Analysis in a PWR." ECN Memo No. 8.904.00-GR17. July 1982.
- l' 3. T.G. Theofanous and Chien-Hsiung Lee, "The Direct Heating Problem," Presentation j to the Containment Loads Working Group Meeting, Rockville, MD, March 1984
- 4. T.G. Theofanous et al., " Decay of Buoyancy Driven Stratified Layers with Applications to PTS " NUREG/CR-3700, May 1984.
- 5. K. Hu, P. Gherson and T.G. Theofanous "The Large Scale Simulation of Debris Bed Coolability," AIChE Symposium Series 236. Vol. 80, pp. 380-384, 1984.
- 6. K. Hu and T.G. Theofanous, " Scale Effects and Structure of Dryout Zone in Debris Bed Coolability Experiments," These Proceedings.
24-10 i
3
?
,/
W FEB 0 91967
/ -FEMORANDUM FOR: Vincent S. Noonan, Project Director i Project Directorate #5 I Division of PWP Licensing-A I d
FROM: Charles E. Rossi, Assistant Director .:
Division of PWR Licensing-A
SUBJECT:
SEABROOK EMERGENCY PLANNING STUDY - 5,'
TREATMENT OF PREEXISTING LEAKS IN CONTAINMENT As a part of the staff evaluation of the applicant's submittal on the Seabrook '
Statien Emergency Planning 7one, the treatment of preexisting leaks regarding the containment isolation dependability was reviewed by the staff in the -
Engineering Branch. Attached is a draft evaluation of the treatr.ent of preexisting leaks in containment.
This evaluation concludes that (1) the Seabrook purge and vent valves in a fully closed configuration should be capable o' withstandina the severe !
accident induced pressure, and (?) the applicant has presented a reasonable approach for considering preexisting leaks in the containrrent.
Or U :;l:.'-.c3by Charles E. Rossi, Assistant Director Division of PWR Licensing-A j
Attachment:
As stated cc: T. Novak R. Ballard S. Long V. Nerses ;
S. Newberry G. Bagchi
Contact:
G. Bagchi '
X27070 l
- _ - _ _ . _ _ _ . . _ , _ _ _ _ _ _ _ _ - - , - - _ _ _ _ - _ _ . - - _ . _ _ _ - _ _ _ _ _ _ - - - _ , _ . - - . ]
Da ATTACHMENT 1 SEARROOK STATION EMERGENCY PLANNING ZONE STilDY -
EVALUATION OF TREATMENT OF -
PPEEXISTING LEAKS IN CONTAINMENT i Rackground: Demonstration of operability of the containment purge Ited vent valves aoainst internal pressure from a design basis accident is required to assure dependability of containment isolation. The safety evaluation of the operability qualification is documented in NIIREG-0896 Supplement Number 5,
' Appendix Q. The sta'f obtained the basic infomation on the Seabrock
- containment purge and vent valves as a part of the licensing review under the TMI Action item II.E.4.7 In order to assess the behavior of these valves in the severe accident environment, the staff has used this basic infomation on the valves, and assumed that the valves would be fully closed during the severe accident phase because of their demonstrated ability to close under the desion basis accident condition.
i Also, based on numerous reports from various licensees on unavailability of containment function and reports of failures of type C leak rate tests of containment isolation valves, it is important in any risk analysis to take into account the effect of preexisting leaks that may have gone undetected
! during the plant operation prior to a postulated severe accident. Therefore, a study of unavailability of containments was undertaken under NRC sponsorship by the Pacific Northwest Laboratory (PNL). PNL reported the findings of its
- study in NUREG/CR 4220. " Reliability Analysis of Containment Isolation
- Systems." This study estimates the probability of larger leaks (28 scuare inchesl to be in the range of 0.001 to 0.01 with a point estimate of 0.005.
During its review of the Seabrook Emergency Planning Zone (EPZ) study, the i staff requested additional infomation from the applicant to address the effect t
of preexisting leaks in its assessment of the probability of various release categories.
The purpose of this evaluation is to (1) document the staff assessment of the capability of the Seabrook purge and vent valves to resist the severe accident environment and (?) to detemine the reasonableness of the applicant's approach for the consideration of preexisting containment leaks.
i " Evaluation:
- As reported in NUREG-0896 Supplement Number 5, the Seabrook purge
!. and vent valves are 8 inch butterfly type Posi-Seal (Model 78988), Class 150 l with Matryx airactuator (Model 76069-SR60). There is a pair of valves in each flow path with independent flow interruption capabilities and on loss of eir
- the valves close due to spring loading. These valve assemblies are analyzed for seismic loading of 3a per axis with loads alone all axes acting simultaneously and superimposed aerodynamic load simulating the pressure load from a design basis accident. The combined stresses are kept under the ASME i
Code allowable values. The valve seat material is resistant to containment spraychemicalsandradiagion. The 1-year accident dose rate is calculated to i
beg approximately 1.? x 10 rads compared to the material resistance sevel of 10 rads. These valves also have screens in elbows upstream of the valves to
- stop debris from entering the valve seating area.
i I
a
- _. ____ _ - ~.__ _ . _ _ _ ____ _ _ _ _ _ _ __ _ _ _
Q -?-
The Posi-Seal E" Class 150 wafer-type butterfly valves have an ANSI rating of 230 psig at 300*F 1.e. a capability to hold against a pressure of 230 psio at a temperature of 300*F. The highest stresses due to the 39 seismic;and combined design basis accident pressure of 60 psig are as follows: ,
Valve stem 23,331 psig (52,500 psig allowable)
Disc pin 21,699 psig (57,500 psig allowablel Rased on the above discussion these valves are capable o' resisting the containment capability pressure of 157 psig and the pressure of 180 psig at 1*-
hoop strain including the temperature associated with the wet containment condition along with the expected radiation exposure.
The applicant in its letter dated October 31, 1986, responded to the staff request for additional information number 22. In the original study (PLG-0300) the applicant quantified preexisting containment leaks at the rate of 0.1% per day for the release category SS, and with all other release categories estimated the effect of containment failures and bypasses including failure to isolate the containment. It is noted in the apolicant's response that the containment purge and vent valves at Seabrook are leak tested every six months or less and their position is checked monthly. Also, manual isolation valves outside containment are position checked every month. Thus the large pre-existing leakage with a probability estimate of 0.01 to 0.001 in NUREG/CR-4220 may not be appropriate for Seabrook.
In spite of the specific differences at the Seabrook Station, the applicant considered the effects of both small and large preexisting leaks in its FPZ study. For the small preexisting leakaae the applicant estimated that a rate of ten times the allowable leakage would yield zero early fatalities and a small contribution to early in,iuries. For a large leakaoe, assumed to be a six inch valve (on 28 square inch hole), with a conditional probability of SE-3 from NUREG/CR-4220, the apolicant estimated the health impacts using an S6W release cateaory. Their estimate, which they believe to be conservative, is attached as Figure-1.
Conclusion:
Based on its review of the information available the staff concludes that the purge and vent valves in a fully closed configuration should provide reliable isolation of the Seabrook containment under severe accident conditions up to the pressures corresponding to 1% hoop strain in the containment.
The staff also concludes that the applicant has presented a reasonable approach for the consideration of preexisting leaks, both small and large.
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- A p reg o,, UNITED STATES 3 .g g -NUCLEAR REGULATORY COMMISSION L j WASHINGTON, D. C. 20555
.t a
\.,,,.* MAR 31987 MEMORANDUM FOR: Charles E. Rossi, Assistant Director Division of FWR Licensing-A THRU: Victor Benaroya, Chief Facilitier Operations Branch Scott Newberry, Section Leader Facilities Operations Branch FROM: Warren Lyon Facilities Operations Branch
SUBJECT:
STEAM GENERATOR TUBE RUPTURE DURING SFVERE ACCIDENTS AT SEABROOK STATION
REFERENCES:
1 Rossi, Charles E., " Steam Generator Tube Rupture Durina Severe Accidents at Seabrook Station - Draft Interim Report", NRC Memorandum for Vincent A. Noonan, Dec. 8, 1986.
P. Theofanous, T. G., " Review Corments on Seabrook Station Steam Generator Tube Response During Severe Accidents (a draft NUREG Report dated 12/1E/86) and Related Sections of Technical Evaluation of the EPZ Sensitivity Study for Seabrook (a draft BNL Report dated 12/5/86), Dept. of Chem and Nuc. Eng., Univ. of Calif., Jan. 17,-1987.
Plant Name: Seabrook Station, Unit 1 Docket Number: 50-443 Resp. Directorate: PWR Directorate #5 l Project Manager: Victor Nerses i Review Branch: Facilities Operations Branch, DPL-A Review Status: Ongoing We previously transmitted a draft assessment of Steam Generator Tube Pupture i at Seabrook Station (Ref.1). We have updated this document and have l included changes identified by Theofanous (Ref. 2). The updated report is enclosed.
This report addresses the state of knowledae pertaining to Steam Generator
. Tube Rupture during postulated severe accidents, and the application of this l knowledge to the Seabrook Station nuclear pcwer plant.
There has been no attempt to comprehensively address all aspects of the subject, and many topics, such as operator actions, are merely identified as areas where further work is indicated.
l
Contact:
W. Lyon j
x28053 l
1 -
Charles E. Rossi The major conclusions are as follows:
- 1. Study of SGTR due to severe accident conditions is difficult due to the complexity of the phenomena and the developmental nature of analysis techniques.
- 2. Further work is necessary to conclude that SGTR is unlikely under conditions associated with a severe accident.
- 3. SGTR dua to severe accident conditions can be shown not to be a problem if the reactor coolant system is depressurized.
'arren Lyon Facilities Operations Branch
Enclosure:
As stated cc: T. Novak S. Long F. Coffman J. Han V. Leung V. Noonan V. Nerses M. Cunningham J. Purphy R. Barrett I
1
. - - = - . - - + n -m.. , w, -- . - ,. -, .,,_v . . - - - . . - , - . - . . ., -w- - - , ,
ENCLOSURE SEABR00V. STATION STEAM GENERATOR TUBE RESPONSE DURING SEVERE ACCIDENTS JANUARY 27, 1987
Foreword This report addresses the state of knowledge pertainino to Steam Generator Tube Rupture during postulated severe accidents (approach to core melt and core melt), and the application of this knowledae to the Seabrook Station nuclear power plant. This is an interim report, prepared with the assumption that the work and assessment will continue. The report does not cover all material received from Public Service of New Hampshire (PSNH) and its contractors, nor is it intended to provide a complete coverage of the issue. It does, however, identify a number of areas where work has been accomplished, it provides an assessment of that work, and it provides suggestions for future work which may be needed to resolve the issue.
Actual resolution effort will depend upon addressing such issues as a pressurized Reactor Coolant System vs. one which has been depressurized.
i
?
Nomenclature AC Alternating current ACRS Advisory Committee on Reactor Safeguards BNL Brookhaven National Laboratory EPRI Electric Power Research Institute ICC Inadequate core cooling KW Kilowatt LM Larson Miller parameter LOCA Loss of coolant accident MW Megawatt NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system PDS Plant damage state (See below)
PORY Pressure operated relief valve PRA Probabilistic Risk Assessment PSNH Public Service of New Hampshire RAI Request for additional information RCP Reactor coolant pump RCS Reactor coolant system RWST Refueling water storage tank SG Steam generator SGTR Steam generator tube rupture Staff The NRC Staff Plant damage states are used to classify conditions as follows:
1 Early core melt, low RCS pressure at time of reactor vessel failure, RWST injection not initiated 2 Early core melt, low RCS pressure at time of reactor vessel failure, RWST injection initiated 3 Early core melt, high RCS pressure at time of reactor vessel failure, RWST injection not initiated 4 Farly core melt, high PCS pressure at time of reactor vessel failure, PWST injection initiated 5 Late core melt, low RCS pressure at time of reactor vessel failure, RWST injection not initiated 6 Late core melt, low PCS pressure at time of reactor vessel failure, RWST injection initiated 7 Late core melt, high RCS pressure at time of reactor vessel failure, RWST injection not initiated 8 Late core melt, high RCS pressure at tine of reactor vessel failure, RWST injection initiated 9 Core melt with non-isolated SGTR 3
__ _ I
A Containment intact at start of core melt, containment heat and fission product removal available B Containment intact at start of core melt, containment heat removal only available C Containment intact at start of core melt, containment fission product removal only available D Containment intact at start of core melt, none of the containment func-tions available E Containment not intact at start of core melt, activity release filtered F Containment not intact at start of core melt, containment opening larger than three inch diameter FP Containment not intact at start of core melt, containment opening smaller than three inch diameter FA Aircraft crash l
l l
4
- 1. OVERVIEW AND
SUMMARY
The Public Service of New Hampshire (PSNH) has presented information to show that the Seabrook Station containment is one of the strongest of any nuclear power plant. It also contains one of the laroest volumes. This combination leads to a conclusion that the containment has the capability to either significantly delay or prevent the release of large quantities of radioactive material.during and following a severe (core damage or core melt) accident.
Based on' this premise, any sionificant risk associated with Seabrook Station will likely be found in accidents which bypass containment.
Recognizing this, the Staff and PSNH have explored containment bypass l
possibilities. One possibility, the topic of this report, and a potential issue that has been under investigation by industry and the Staff for several years, is the loss of steam generator tube integrity due to generation of high temperatures at hioh pressure during a core melt accident. The potential concern-involves movement of high temperature fluid from the region of the melting reactor core into the steam generator tubes, with a resultant over-heating of the tubes which leads to their rupture. High pressure fluid containing radioactive material from the melting core would thereby be released to the secondary side of the steam eenerators, from where it could be released to the environment via the steam generator relief valves, thus bypassing containment.
i For steam generator tube rupture (SGTR) to be a concern as addressed here, one must have a core damage (or melt) cordition in progress with no water on the steam generator secondary side. The principal contributor to this condition is estimated to be a loss of all AC power concurrent with a loss of all turbine driven feedwater to the steam generators. PSNH has investigated the possibility of encountering conditions which can contribute to SGTR and has determined the likelihood to be less than 4 X 10-5 per reactor year. This is sufficiently high, and the potential consequences of SGTR under severe accident conditions are sufficiently great, that further investigation has been necessary. This investigation is ongoing. This report provides an interim assessment of the status of the investigation, as well as a projection of expected results.
5
.. 1 Study of SGTR due to severe accident conditions is difficult. The phenomena are complex, and most analysis techniques used to investigate nuclear power plant behavior have utilized assumptions which are not applicable here. The principal complication is the multidimensional character of fluid behavior in the reactor coolant system. Suitable computer programs are just beginning to become available. Suitable experimental information is just being developed.
Hence, pioneering work, such as provided by PSNH in investigation of this issue, can be expected to have weaknesses as well as strengths. We have found this expectation to be true.
The work reported by PSNH and its contractors is highly informative and addresses most aspects of the SGTR issue. It is based upon knowledge of what takes place within the Nuclear Steam Supply System (NSSS), upon a major computer program that is under development and is beino verified (MAAP), and upon information derived from an experimental program at Westinghouse. The following is a summary of the reported information and our assessment:
- 1. Mathematical modelino. Expected phenomena, experimental infomation pertinent to the phenomena, and modeling assumptions have been addressed for each of the major components of the NSSS which are affected. Mul ti-dimensional fluid flow and energy transport have been established as dominant over most of the conditions of interest. We consider this crea to be in a preliminary staae of development, and there are some potential difficulties, which include:
- a. Certain modeling assumptions are overly optimistic. An example is the assumption of complete mixing in the steam generator inlet plenum which tends to reduce the temperature of fluid entering the steam generator tubes. This assumption is not supported by the available experimental evidence, and the effects of the assumption are not balanced by identifiable pessimistic assumptions elsewhere in the analysis,
- b. Experimental evidence is preliminary. The experimental facility at Westinghouse is providing information pertinent to this issue.
However, testing has been limited to conditions which are only 6
4 rouchly scaled to NSSS representation. This is due to a logical progression in the test planning and facility development. -Data from apparently well scaled test conditions are just becoming available.
No other test facility addresses certain aspects of this issue.
- c. The computer program used as the basis for much of the work has not been verified, nor is documentation available. We understand a
, verification _ program and an effort to provide documentation are underway. (PSNH contractors have offered to discuss this information with us. Our review has not progressed to the stage where we can make use of this offer.) Although the phenomena we understand to be 4 modeled by the code appear adequate for the purposes needed here, and the code results appear reasonable subject to our concerns as expressed elsewhere in this report, this is not sufficient infonnation to accept the analysis results.
- 2. Seabrook Station Representation. The basic analyses and sensitivity studies have been based upon a plant configuration in which the NSSS state is assumed. Most of the assumed state conditions are reasonable. There are exceptions. For example, the steam generator secondary side is
- assumed to be at a pressure corresponding to secondary side relief valve settings, and creep rupture of tubes is reported for this state. The 4
, resulting conclusions are similarly based upo.1 this state. We believe there is sufficient likelihood the secondary side will be depressurized j that this case should be considered. Depressurization would roughly
- double tube stress since the secondary side pressure would be decreased from roughly 1100 psi to atmospheric pressure while the RCS pressure remained at approximately 2300 psi.
- 3. Sensitivity Studies. PSNH and its contractors have performed a wide ranging sensitivity study as part of an assessment of the impact of various modeling assumptions and the state of the plant. Although this yields valuable information and insight, sensitivity studies should be l
7
t approached with caution. They are only as good as the basic modeling.
< The impact of our difficulty with assumptions such as the behavior of the steam generator inlet plenum is not addressed in the sensitivity study, and could impact the results and conclusions. ['
t
- 4. Operator Actions. Plant response can be drastically altered by operator ,
actions durino a severe accident.- SGTR is no exception. A number of operator responses have been discussed with PSNH. Although many of these were postulated actions, significant information has been ' developed from these postulations. Recognition that operator actions ccUld depressurize the steam generator secondary side is one item raised ddring the review.
Depressurization of the reactor coolant system via the pressurizer Pressure Operated Relief Valve (PORV) to avoid the SGTR problem is another. \ .1 s
We find that the topic of SGTR is in a developing state, with knowledge being rapidly accumulated. Further work is necessary to conclude that SGTR is' unlikely under all conditions associated with a severe accident. o 1
Existing knowledge can be used to support a conclusion that SGTR is not ti problem if the RCS is depressurized. Consequently, reasonable assurance that progressions toward core melt would not occur at high RCS pressure, coupled with supporting evidence in regard to steam generator tube response, would alleviate our concern regarding SGTR under severe accident conditions. We have not conducted an evaluation of the trade-offs c sociated with such an approach, nor have we been provided with information that would either support i
or negate RCS depressurization under severe accident conditions. We have not provided a recommendation regarding whether RCS depressurization is attractive when all pertinent factors are considered.
Our judgement is that a carefully conducted thorouah evaluation on the part of PSNH can establish that the likelihood that a SGTR will result due to overheating during severe accidents which initiate from power operation is sufficiently small that the risk associated with this event can be shown to be negligible. Our judgement is preliminary and has not been substantiated.
Theofanous (Ref. 22) believes depressurization should be accomplished, and does not foresee any significant reasons why this should not be done.
8 L _
Determination of the correctness of a judgement regarding SGTR under severe accident conditions originating from power operation with the RCS at high pressure can be based upon a coirbination of analytic and experimenti investi-gations. The ongoing test at Westinghouse in which reasonably close similitude is claimed between the test facility and appropriate parts of a Westinohouse four loop NSSS will provide key data which can be applied to assist in the development and confirmation of analysis techniques. Ilse of selected test data from other facilities and further examination of the analysis techniques, coupled with necessary changes when they are uncovered, should provide sufficient confirmation that reasonable reliance can be placed upon accident analyses pertinent to this issue. Suitable analyses can then provide a sufficient foundation to resolve this issue. Such a program will represent a formidable undertaking.
4 Theofanous (Ref. 22) states this judgement to be "... an overly optimistic and inappropriate .iudgement for the Regulatory to make at this time". He continues with "...the procedure outlined to ' substantiate this judgement' is unrealistic and incomplete". Although we continue to believe the issue can eventually be established to not contribute significantly to risk, we certainly agree with Theofonous' assessment that such a determination will not be easy.
l Further, our " judgement" is preliminary and unsubstentiated, and is not to be l used as the basis for any regulatory findings until reasonably established to be incorrect or correct.
9
- 2. INTRODUCTION The Public Service of New Hampshire (PSNH) reporting of Seabrook response to accident conditions in References 1 - 4 represents one of the most comprehen-sive investigations of nuclear power plant accidents in a specific plant that we have encountered. Some accidents which have a significant impact upon risk are treated more comprehensively than previously reported by any investigator.
For example, References 3 and a describe an investigation of LOCA outside of containment that is more comprehensive than any we have reviewed. Many of the commonly used conservatisms, which distort the perception of accident impact, have been removed. What results is a serious attempt to better represent plant response to severe accident conditions, with particular attention to items which have previously been identified as having a serious impact upon risk.
PSNH has presented information to show that Seabrook Station has one of the strongest containments of any nuclear power plant. It is also one of the largest with respect to containment volume. The combination of large volume and strength leads PSNH to a conclusion that the containment can mitigate virtually every severe accident and, at the worst, can significantly delay release of meaningful quantities of radioactive material during and folloving core melt accidents. Most core melt accidents can be contained within the Seabrook Station contair-n-nt, and, if this is accomplished, little radioactive material will escape. The fu:, mitigative capability of the Seabrook contain-ment will be realized if there are no " holes" in the containment. Such holes can exist if any of the following occur:
- 1. Contairment is not properly closed (isolated), such as can occur if containment ventilation is not properly closed upon receipt of a contain-ment isolation signal,
- 2. A failure occurs which allows the containment atmosphere to escape, such as failure of a containment penetration due to a combination of high pressure and high temperature, or 10
- 3. A failure occurs which allows material to move directly from the h'uclear Steam Supply System (NSSS), principally the Reactor Coolant System (RCS),
to the environment, such as occurs with the traditional " Event V" (Ref.
5), with LOCA outside containment leading to core melt and the release of I radioactive material via the LOCA flow pathway.
Clearly, if PSNH conclusions regarding containment strength are verified, there will be little risk associated with accidents at Seabrook Station unless containment is bypassed. Therefore, core damage accidents with containment l hypass deserve careful attention. PSNH has reported studying some bypass accidents in detail (Pefs. 3, 4, 12, 17, and 18). Such studies have led them to conclude that certain bypass accidents at Seabrook, such as LOCA outside containment, engender significantly less risk than previously believed. Other bypass accidents have only recently been identified and accident investigation is not complete.
One potential area for bypass, as identified above, involves paths between the 1 RCS and th' e environment. Certain phenomena can potentially lead to such paths.
These involve multidimensional fluid behavior and fission product transport within the RCS during the approach to core melt and during the core melt process. Consideration of these phenomena has a significant impact upon RCS response, including potentially the location of RCS failure. There are many possible implications, including the possibility that the impact of RCS failure f f on containment may have been overestimated in past analyses. The implication of interest here is that failure to accurately model RCS fluid and fission product heating behavior might result in an RCS failure which bypasses con-tainment. The only area discovered where this is of immediate concern
- involves the Steam Generator (SG) tubes. If these fail during a core melt accident while the RCS is at high pressure, there is a high potential of a major release via the SG relief valves or the SG Pressure Operated Relief Valves-(PORVs), which vent directly to the environment; or via a rupture in a l- steam line outside containment.
l The general concern addressed in this report is the rupture of multiple SG tubes in response to high temperature, which in turn is a result of core 11
l l
uncovery. This accident sequence should be of concern any time there is a core melt with the RCS at high pressure in combination with no water in the SG secondary sides. These conditions lead to a potential for natural circulation transport phenomena to significantly heat the tubes prior to breach of the reactor vessel. If this occurs, the resulting loss of tube strength could lead to tube rupture. If tube rupture occurs, and any of the secondary side valves are open, the secondary side is breached outside containment. Alternatively, if the RCS pressure is above the SG relief valve setpoints, containment is similarly bypassed. There is no substantiation which establishes that these valves will close after being exposed to such an environment, nor has it been established that other secondary side failures will not occur. This area has not been adequately investigated, and is not recognized as a release path in the early Pickard, Lowe and Garrick work on risk investigation at Seabrook Station (Refs.1-4), nor is it addressed in any of the other PRAs we have received. It has been addressed in more recent work (Refs. 12, 17, and 18).
The concern was expressed as the rupture of multiple steam generator tubes. We do not believe single tube ruptures will occur under the severe accident conditions of interest. The reason for this is that if one tube ruptures, or even begins to leak significantly, this will induce flow of hot RCS fluid toward the leak. Therefore, the location of tube rupture will probably ouickly become hotter. If high temperature is what led to the break, a higher temperature can only make it worse. Tubes in the vicinity of the break will be exposed to the high velocity break flow, in additional to high temperature, weakening them and, we believe, ouickly leading to their failure. We believe this cascading effect would rapidly propaaate to multiple tube rupture, stopping only when sufficient RCS depressurization has occurred that tubes are no longer stressed by a significant pressure differential across their walls.
This belief is based in part on the assumption that SG tube degradation
- has been controlled and there are no " outliers" which fail significantly sooner than other tubes due to existing tube imperfections.
I I?
Although this report is limited to SG tube rupture, there are other SG compo-nents which separate PCS fluid from the SG secondary side. These components, such as the SG tube sheet, must be investicated to achieve completer.ess in the investigation of containment bypass via the steam generator.
An initial consideration in investigation of the SG tube rupture issue is "What is the likelihood of attaining conditions where SG tube response could be of concern?" Principally, the conditions are loss of all SG feedwater with a simultaneous loss of PCS makeup capability; conditions which result, for example, from a loss of all AC electrical power with the simultaneous loss of the turbine driven auxiliary feedwater pump. PSNH estimated this condition to ~
have a mean annual frequency of less than 4.5 X 10-5 per reactor year ** (Ref.
17). A value of this magnitude is sufficiently hich that tube response must be considered.
l We have not fully investigated this value or its uncertainty and consequently are not verifying it as " correct" via its usage here. We do believe it is of a reasonable magnitude, and as such, that further work on SG tube rupture is indicated.
13
- 3. STEAM GENERATOR TUBE RUPTURE (SGTR1 UNDER SEVERE ACCIDENT CONDITIONS 3.1. Description of Phenomena and Potential Concern.
The RCS is generally modeled with a one dimensional representation of fluid flow, and in some cases with parallel one dimensional modeling in regions such as the reactor vessel. This has been particularly true for PRAs, where to our knowledge, all have been based upon computer code analyses which incorporated single dimensional representations of fluid behavior within the RCS. Addi-tionally, movement of the source of heat due to fission product migration is seldom modeled.
The possibility of RCS behavior being different from what is generally repre-sented during severe accidents has been recognized for some time. Winters (Ref. 6) identified aspects of the problem in 1982. Denny identified poten-tially important aspects of natural circulation, and Denny and Sehgal (Ref. 71 l
provided preliminary multidimensional analysis results in 1983. The topic was discussed by an NRC containment response working group and with the ACRS (Refs.
19 and 20), it was the subject of an NRC/ Industry meeting (Ref. 8) and a formal request for work within NRC (Ref. 9), and SGTR possibilities were identified (Ref. 21) in 1984. Potential impact upon SGTR was estimated on a preliminary basis (Ref. 10), and experimental data were presented from an ongoing series of l tests (Ref.11), in 1985. Numerous analysis results have been published since the early publications of Denny and Sehgal which represent work sponsored by both industry and the NRC. Powever, there is no published analysis of overall NSSS response to a broad range of severe accident conditions which includes these phenomena, and which is based upon accident analysis methods which have been subjected to broad peer review and acceptance. This introduces a difficulty into review of SGTR during severe accidents with respect to the impact upon the Seabrook Station risk evaluation. As will be seen, sufficient work has been accomplished that what appear to be reasonable conclusions can be formulated, although confirmation will require additional effort. As will further be seen, there appear to be operational methods which can negate the problem, although the impact on other a:,vects of plant operation has not been evaluated.
14
The potential misrepresentation of system response of concern here ster.s from the fluid flow behavior inherent in one dimensional modeling as utilized by most accident analysis codes. Such modeling typically represents flow through ,
the reactor core as determined by the water boiloff rate from the lower core or lower plenum. This rate becomes small as the water level approaches the bottom of the core. Typical calculations (see historical references which were previously discussed) indicate that the flow rate due to natural convection which occurs in a multidimensional manner is of the order of ten or more times that of the flow due to boiloff. Hence, the calculations are typically based on a minor contributor to flow, and the major contributor is neglected.
The modeling difficulty also applies to upper plenum behavior. One dimensional modeling of any fluid (liquid, vapor, or pas) that passes through the core is typically assumed to flow through the upper plenum and out the hot leg.
This modeling is incorrect under severe accident conditions where a major portion of the core has been uncovered or the core is being vapor or gas cooled since strong recirculation patterns will develop which thermally link the core and upper plenum. At pressures in the range of 2250 psi, the linkage is strong, and some of the upper plenum component temperatures can be expected to closely follow core temperature during the early stages of the approach to core melt. The strength of the linkage diminishes with decreasing pressure.
Information also exists which illustrates a decrease in linkage with increasing hydrogen concentration and core damage (although initial production of hydrogen may enhance circulation due to the buoyant gas " pushing" its way toward upper regions of the reactor vessel).
Correct consideration of the hot leg and steam generator behavior leads to calculation of significantly different behavior when contrasted to one dimensional modeling. Hot fluid, at a temperature far greater than predicted via a one dimensional model, will enter the upper portion of the hot legs from the reactor vessel, and flow toward the inlet plenum of the steam generators.
Displaced colder fluid will return to the reactor vessel upper plenum along the bottom of the hot legs. Circulatory patterns will becere established in the steam generator inlet plena in which some of the hot incoming fluid is mixed 15
with plenum fluid. Fluid from the steam generator inlet plena will flow into some of the steam generator tubes in the nominal forward direction, displacing fluid in the steam generator outlet plena. This displaced fluid will flow throuah other tubes in a nominal reverse direction, reentering the steam generator inlet plena. (All of these flows have been observed experimentally as described in References 11, 13, and 14). This mechanism has the potential to transport hot fluid from the reactor vessel into the steam generator tubes during core heatup and melt, with the result of creating the potential of overheating the tubes if there is no water on the steam generator secondary side.
There are other possibilities which could challenge tube integrity as well.
For example, RCP seal LOCA or a smal1~ RCS cold leo break introduce a low pressure region between two regions where a liquid seal or plug may exist -the crossover pipe between the RCP and the SG, and the lower reactor vessel. Under approach to core melt conditions, one path for flow is through the SG tubes, through the crossover pipe seal, and cut the break. (Note this does not remove the seal - the steam simply bubbles through it). This flow path of hot steam through SG tubes and the associated thermal impact on the tubes must be considered. Another tube challenge can result due to emergency procedures.
Many plant Inadeauate Core Cooling (ICC) emergency procedures specify RCP l
operation if conditions exist which indicate an approach to core melt, and alternate mitigative measures have failed. Such a step could circulate hot fluid through the RCS, including the tubes. Although this may slightly extend the time to core melt, it may be an unattractive approach if it also introduces a high likelihood of loss of tube integrity. To our knowledge, these contrasting responses and the impact upon risk have not been studied. (Note the likelihood of encountering the emergency procedures problem situation is small, but it does exist.)
A final phenomenon that has received inadequate attention during conditions leading to core melt is fission product movement. Typical one dimension accident code calculations take such movement into account from the viewpoint of radiological hazard, but do not include the influence upon he6t ceneration.
Approximately a quarter of the heat producino radioisotopes probably has left 16 -
the core under the conditions of interest, and substantial deposits can be expected in the upper plenum structure. This could have a sionificant influ-ence upon thermal response, particularly if some of this material leaves the reactor vessel and enters the hot legs.
As will be seen in the following sections, PSNH has addressed many of these issues in the most comprehensive study of this problem that we have encountered.
3.?. Seabrook Station Steam Generator Integrity 3.2.1 Issues Addressed By PSNH The PSNH has addressed many of the issues applicable to SG tube response to severe accident conditions (Refs. 12, 17, and 18). Analysis results were summarized which were intended to determine the thermal response of SG tubes under severe accident conditions. Basic analysis assumptions pertinent to the state of the plant were:
- 1. The steam generators must be dry to experience a significant thermal transient since, if the SG secondary side contains water, the tubes cannot overheat.
- 2. Station blackout conditions (Loss of all AC power) exist.
Analyses were conducted for the following:
- 3. Station blackout with operator actions 4 Uncertainty evaluation 17
Possible operator actions considered included:
- 1. Start steam turbine driven auxiliary feed water flow
- 2. Restore emergency AC power (diesels and/or switchgear)
- 3. Shed nonessential loads 4 Open RCS POPVs when core exit temperatures exceed 1200 F.
A number of other operator actions one might expect were discussed during a meeting with the PSNH at BNL on October 17, 1986, including:
- 5. SG blowdown and depressuri7ation to enable filline the SGs by the condensate booster pumps or from fire water systems. (There are two diesel driven pumps and one electrically driven pump at Seabrook Station. The ability to use these for in,iection into the SGs has not been confirmed.)
- 6. RCP operation, a step that is not possible unless off site electrical power has been restored. (PSNH felt the likelihood was sufficiently low that there would be negligible effect on risk.)
3.2.2 Likelihood of Conditions Leading to Tube Failure PSNH addressed the question of conditions necessary for SGTR in the response to the Staff Recuest for Additional Information (RAll 47 (Ref.
17). In this response, PSNH stated the risk to be small for the following reasons:
- 1. The frequency of high pressure core melt with dry steam generators is very small.
- 2. Given the postulated occurrence of a high pressure core melt with dry steam generators, creep rupture of the SG tubes is not a credible failure mode.
18
- 3. A large number of tubes must fail to produce an early large contain-ment bypass.
- 4. All three of the following must octnr in order for there to be a containment bypass:
- a. Failure to recover water to the SG
- b. Failure to depressurize the RCS
- c. SG tube creep failure
.x 3.2.3 PORV Considerations PORV operation as identified in item A, above, is not specifically contained in Seabrook Station emergency procedures, but is belf tved by PSNH to be a logical operator response as an attempt to depressurire and obtain water from the accumulators. (Operator monitoring of the tempera-tures is specifically identified in the procedures for loss of all AC powerconditions.) In addition to potential core cooling via the accumu-lator water, opening the FORVs is claimed to have the following effects:
- 1. It reduces stresses in all primary system components
- 2. FORY flow overrides natural circulation such that hiah fluid temper-atures are not attained in the SGs, including the tubes.
In response to a staff question, PSNH indicated that the likelihood of bei~ng able to open the PORVs under loss of AC and ICC conditions was high.
They also indicated that one PORY was sufficient since its " worth" is about 50 MW of energy removal in the form of steam, and have presented blowdown rate information in Reference 18. (Note Seabrook is equipped with two PORVs.)
Although we consider the EPRI funded Westinghouse tests pertinent to this issue to be somewhat preliminary with respect to scaling to NSSS condi-tions, some interesting effects have been observed that are worth noting which pertain to PORV operation. These include:
19 j
- 1. Natural circulation flow restores itself readily to the pre-openina condition in the hot leas, core, and consnunication paths between the upper plenum and the upper head following PORV closure.
- 2. Heat transfer in steam generators between the primary and secondary side fluids increases 50% to 75% with periodic venting,
-3. The core is little affected except for the boundary with the hot leg that connects to the pressurizer surge line.
Item 2 is of particular interest since it carries an implication that flow in the steam generator tubes is enhanced by PORY operation (as well as by opening and closing of RCS safety valves). Hence, if one visualizes opening and closing a pressurizer PORV when degraded conditions are well established with the steam generator secondary side depressurized, there may be a tendency to enhance flow of hot RCS fluid through the tubes, with l the potential of causing tube rupture.
l 3.2.4 Loop Seals l Loss of RCS inventory under natural circulation conditions (RCPs not running) is expected to leave the RCS in a condition where water is trapped at low elevations. According to a number of preliminary analyses, such loop water seals or plugs exist at the cross over leg between the SG exit and the RCP inlet, and in the lower region of the reactor pressure vess'el. The absence of these water seals could significantly change circulatory conditions during ICC conditions, with the potential for changing SG tube response. Although we expect a careful examination of behavior in the Seabrook RCS would establish that the seals will remain under most boil down conditions, this expectation needs to be substantiated by suitable analyses which address the range of conditions which can exist during severe accidents.
l Complete loss of the RCS liquid inventory with the RCPs running, followed l by loss of the RCPs, could result in a homogeneous fluid condition in the RCS. Under this condition, fluid heated in the core would flow into the 20
upper plenum, through the hot legs, the steam generators, the RCPs, and back into the reactor vessel and the core via the cold legs. Althouah
, multidimensional fluid flow conditions probably exist in the reactor vessel after RCPs are lost, one may estimate that thermal response is still reasonably realistic if modeling is restricted to one dimension provided the natural convection flow rates are high. For this case, existing analysis codes could be applied to roughly estimate steam generator tube response. If the response was not clear, then multidimen-sional analyses could be applied to estimate the influence. In such a case, uncertainty in the multidimensional analyses might not be of as great a concern as for the situation c7 multidimensional behavior domi-
- nating system response. However, none.<istence of the loop seal due to continuous RCP operation is an unlikel/ situation since the ma,iority of conditions during which steam generator tube integrity is of concern will involve loss of off site AC power, and RCPs will be unavailable. To cur knowledge, a complete, accurate, analysis of a four loop Westinghouse NSSS .
has not been performed for these conditions. In additional to an analysis approach, closure of consideration of this aspect of SG tube behavior could be obtained if the probability of occurrence of the RCS homogeneous fluid condition was established as negligibly small in contrast to other situations where SG tubes were shown to lose integrity, or if the risk
, associated with the condition was established as negligible when compared to other Seabrook Station risks.
A second situation involving free circulation in the RCS might be obtained if one considers the RCPs as being restarted in response to high core temperatures, as prescribed in the emergency procedures. For this case, sufficient head might be developed to clear the loop seals of water, and rehomogenize the RCS fluid, thereby generating the condition described in the previous paragraph. To our knowledge, rehomogenization under these conditions has not been established to occur at Seabrook. Insofar as SGTR at Seabrook is concerned, the issue can be dealt with as outlined in the previous paragraph.
l i
21 f
l
,,.-_y -
- . - _ . . _ . . . _ _ _ _ , - . . , , . _ , . ~ - . . , . . , _ , - - .
l A third situation of removal of loop seals also potentially exists during boil down of the RCS inventory. One may postulate that the ICC condition occurs with the loop seals in place, and that some other mechanism causes l their disruption. This could occur if a sufficient pressure difference occurred across the seals that they were forced out of the low reofons or if superheated stearn passes through the water, thus evaporating it. Several analyses have been conducted which include consideration of some of this behavior, and none showed loss of the seals. To our knowledge, these analyses have not carefully considered the evaporation cuestion or the impact of a sudden pressure surge due to core slump into water in the lower plenum. One would expect that consideration of this condition could be closed if analyses applicable to Seabrook could reasonably establish that the seals remain.
Another condition can be visualized if one considers a LOCA to have occurred in the RCS. For example, a small cold leg LOCA for an RCP seal LOCA) could be located between the two natural seal regions of the crossover leg and the reactor vessel lower plenum. Removal of RCS mass might occur under conditions such that the seal water was evaporated from the crossover leg due to forcing superheated steam through the seal water.
An important aspect of-seal behavior to consider here is that one does not have to empty the crossover leg of water to pass steam through the SG tubes. It is sufficient to bubble steam throuch the seal water. Elimin-ation of consideration of this effect with respect to impact upon risk could be considered on the basis of a thermal-hydraulic investigation of RCS behavior, establishing that the potential impact on risk of the behavior is negligible in comparison to other established risk contributors, or both.
3.2.5 PSNH Modeling Considerations The PSNH has reported application of the MAAP 3.0 code to investigation of natural circulation flow in Seabrook (Refs. 12 and 17). This code treats the major phenomena, including approximations of multidimensional flow and fission product (heating) movement, and is applied to the regions o# the RCS which are affected by the SGTR issue.
22
. Quasi-steady momentum balances and continuity equations are used to represent natural circulation flow, and the steam generator inlet plenum behavior is represented by quasi-steady mixing models. The modeling represents gas and wall temperatures using conventional lumped parameter
! models, with 15 gas control volumes and 17 two dimensional' heat sinks.
(Several volumes are subdivided into further volumes for some types of calculations. The core, for example, contains 70 nodes which comprise the
{ core volume node.) The control volumes are based upon approximations-of the flow patterns which were seen in the Westinghouse experiments on a
- scaled NSSS (Refs; 11, 13, and 14). This basis for definition of control volumes means that deviations from the assumed flow pattern and flow instabilities may not be represer.ted in the model. Experimental evidence i shows that there are asynnetric flow patterns, for example, which are not modeled, and which could lead to tube heating conditions which would not be calculated. Further, although instabilities have not been experimen-tally observed at the Westinghouse test facility, one must accept this evidence with care since testing with fluid conditions which closely simulate those expected in an NSSS are just being initiated.
l Use of the-lumped parameter model recuires further discussion. Unlike i computer codes such as COMMIX, which can detemine flow patterns within
( certain bounds provided the configuration is properly modeled, a lurped parameter model is based more strongly upon a presupposed flow behavior.
Although such representation can be valuable and accurate under certain conditions, such assumed behavior must be verified before it can be accepted. The preliminary Westinghouse experiments, as discussed briefly in the next section of this report, and some COMMIX and MELPROG calcula-tions (Refs.15 and 16), represent steps in this direction, but further l evidence is necessary before we can accept the assumption as verified.
l (The experiments are somewhat preliminary, and the COMMIX and MELPROG calculations have not, to our knowledge, been carefully checked against experimental evidence.) We further note that, to our knowledge, there has I
l l
23
been no independent study of the version of the MAAP code used for the analyses. At a minimum, we believe a reasonable knowledge of code modeling and logic, in addition to a verification program, are necessary
, for acceptance of the calculated results. (We note that EPRI has a PAAP verification program underway.)
, One aspect of the modeling appears worthy of further consideration. The steam generator inlet and outlet plena are anumed to be completely nixed in the PSNH studies being reviewed here, and they are represented by single nodes with uniform properties. The Westinghouse facility test data indicate a partially stratified, partially mixed SG inlet plenum (Ref.
14), and modeling for the test facility is based upon a quasi-steady state model in which partial mixing is assumed at various (limited) locations between streams of different origins. Reference 14 describes the situation as follows:
"The flow in from the hot leg rises rapidly in a plume in the inlet plenum and induces mixing. Some of the cold return flow from the tube bundle does avoid mixing, particularly near the divider which ie furthest from the hot leg. Much of the cold return tubes' flow plunges through the hotter stratified fluid layer that spreads across the bottom of the tube sheet. The mixing flows could be observed from dye injection and from observation of light through the density gradients that resulted. Temperature measurements in the inlet plenum are indicative of mixing. The tubes carrying hot fluid from the inlet plenum were generally concentrated in the area above the hot leg entrance and scattered in the regions further away. Cold
> return tubes were also scattered and were found in the area above the hot leg inlet also."
Test facility modeling of the phenomena uses a six equation approximation 4
which contains an experimentally determined mixing parameter.
We believe the assumption of complete mixing used for the PSNH investiga-tions will reduce SG tube temperatures when contrasted to the experimen-tally identified situation. This modeling and its implications need y
k
further consideration. (This comment is repeated a number of times in the discussion of calculated NSSS response in the following sections of this report.)
3.2.6 Comparisons of Calculations to Experimental Data Several comparisons between MAAP code calculations and experimental data have been briefly described by PSNH and its contractors to the BNL and NRC staffs (Refs. 12, 17, and 18). These are discussed below.
- 1. Core and upper plenum flow rates. The following comparison of experimental and calculated values was presented:
l l
Test Condition Experimental Calculated Flow Rate Flow Rate 28 KW Water Test 0.5a 0.50 0.9 KW SF 6 Test 0.016 0.017
- 2. Hot leg and steam generator natural circulation. Comparison of several parameters was provided:
Calculated Values for Indicated Number of Steam Generator Tubes Experimental Carrying Flow in the Out Direction Item Value 6 12 24 Heat Transfer Rate, KW 2.43 2.0 2.6 2.9 Entering Fluid, OC 30 30.7 29.2 28.4 Exiting Fluid, OC 19 24.2 21.7 18.8 Coolant, OC 10 - 11 9.4 11.2 12.8 where the entering fluid is flowing into the steam generator inlet plenum from the upper portion of the simulated hot leg, and the exiting fluid is flowing from the lower portion of the steam generator inlet plenum back 25
~
toward the simulated reactor vessel along the botton of the hot leg. The coolant temperature is that of the water leaving the secondary side of the simulated steam generator, and thus, can be related to the heat transfer rate from the primary to the secondary sides.
These results are clearly promising. Continuation of the comparisons with a wide range of experimeretal conditions in the same test facility, and with no changes in the modeling exceo'c for the change of experimental conditions and fluid properties, would be helpful in code verification.
Extension of the same modeling approach to other experimental data (such as flow in ducts and components) would provide further confirmation.
Completion of confirmation of modeling adequacy could typically include comparisons of existing data obtained in large facilities, selected contrasting of alternate calculational methods to portions of the code under consideration here, and establishment that scaling is adequately 3 e represented by the code.
3.2.7 Calculated Seabrook Thermal Response to Severe Accidents Calculated behavior to selected accident conditions has been summarized by PSNH. Principal results and our comments are as follows:
- 1. Peak Steam Generator Temperature for loss of AC Power and Loss of Feed Water Flow. The following temperatures and flow rates were calcu-l lated at the indicated condition:
! Location Temperature, K Flow Pate, kg/sec Core (Peak) 1800 18 (recirculating between Upper Plenum 1160 upper plenum and core)
Hot Leg 760 (wall) 2.4 (countercurrent)
SG Inlet Plenum 850 -
SG Tube 700 (wall maximum) 3.3 (total in each direction)
SG Outlet Plenum 640 -
26
PSNH indicated that the hottest core node would melt at about 30 seconds from the time of these values, and that the generated hydrogen and blockage due to relocated core material would cause natural circulation between the core and the upper plenum to almost stop. At this point, the upper plenum would begin to cool due to energy transfer to the hot legs.
Plys (Ref. 181 presents additional information which shows temperatures continue to increase after vessel blowdown, with the peak upper plenum 0
temperature exceeding 1200 K for a short time. The tube temperature continues to increase for the time of the calculation (20,000 sec, with vessel rupture at 11,600 sec), reaching a maximum of about 1020 K. We would be interested in seeing plots of other parameters over the span of the calculations, including the hot leg and SG plena temperatures, to better understand the interactions and modeling.
In response to a question, PSNH indicated they had not performed a detailed analysis of reactor vessel hot leg nozzle thermal behavior, but felt a temperature of the order of 1000 K was necessary to cause failure.
Discussion also identified that there was significant steam circulatory flow in the secondary side of the steam generator tubes, and that this steam, which was at a pressure corresponding to the steam generator safety valve settings, represented a significant heat sink. Further, it was an effective medium for transferring heat from hot tubes to colder tubes, thus tending to reduce the maximum tube temperature. This raises a question of what results would be obtained if the steam generators were depressurized to atmospheric pressure, thus maximizing pressure differen-tial across the tubes and simultaneously removing a heat sink which could influence temperatures throughout the NSSS. (A sensitivity analysis was conducted in which this was one of the parameters.)
Information presented in Reference 12 and the above summary table shows fluid flow rates in the hot leg of roughly 2 kg/sec as contrasted with a rate above 3 kg/sec in the SG tubes for the time after effective boiloff of water from the core until melt through of the reactor vessel. Cooling 27 l
l via steam contained in the SG secondary side is thus an effective medium for cooling the SG inlet plenum. The total mixing assumption pertinent to fluid in the plenum is, in turn, effective in preventing hot fluid from ,
reaching the tubes. This high tube flow rate is also effective in )
transferring heat from the reactor vessel to the SG seconda*y side, thus helping to limit fluid temperature in the hot legs as well.
We believe a study would be beneficial of behavior with the SG secondary side depressurized after SG dry out. Now there would be no heat sink on the secondary side, and tube flow rates may be lower due to less of a driving force for natural convection flow in the SG. Further, we would expect to see further stratification in both the hot leg and the SG inlet plenum (the latter not being allowed in the PSNP supported analyses due to the modeling assumption of complete mixing). We pose the question of whether temperatures may be significantly above what was calculated by PSNP and its contractors under these conditions.
- 2. Operator Induced Depressurization. This calculation was based on the assumption that the operator would open an RCS PORY when the core exit thermocouples indicated 12000F. The calculations indicated accumulator discharge approximately 1400 sec after opening the PORV, with the RCS depressurized prior to vessel failure. The accumulators were emptied at about 10,600 sec, and vessel failure occurred 2000 sec later. Accumulator I
water was found to cause a small additional amount of hydrogen production.
Phenomena associated with depressurization and hydrogen decreased the effectiveness of heat transfer between the core and other reaions of the l
NSSS. Steam generator inlet plenum temperature reached a peak of roughly I
8500 K during the depressurization, then cooled, and remained below 650 0K for the remainder of the calculation (20,000 see total calculation time, with PORY opening at approximately 8000 sec). Maximum tube temperature was about 6500 K, and was reached at 20,000 sec, being identical to the SG inlet plenum temperature at that time. (Note RCS pressure is that of the containment following depressurization earlier in the calculation.)
28 l _ _ . _ _ _
We note that RCS pressure behavior (Ref.18, Figure 4-4) is different for the base case and the PORY opening case prior to the time of opening of the PORV. We would like to discuss these differences for all parameters and we would like to understand the reasons they exist. (We note there is little difference in temperature over the range in question, and temperature is the important parameter for the SGTR issue.)
Volatile fission products represent about 20% of the decay heat, and the behavior of this energy source is calculated in the MAAP code. The calculations illustrated movement of the decay heat source. About 10% of the decay heat was associated with fission pmducts which were in the upper plenum at the time of vessel failure. A small amount was in the het legs, as was also the case for the pressurizer. The amount in the steam generator tubes was not significant. (Most of the CsI was in the upper plenum at the time of vessel failure, with about 10% of the Csl in the bot
' legs.)
l l
- 3. Other Variations and Uncertainty. Several sensitivity calculations were performed to obtain a better understanding of behavior. These j included:
l l a. Higher core melt temperature l b. RCP seal failure l c. SG secondary side blowdown
- d. Core resistance variation
! e. Reduced SG tube circulation 1
- f. Core blockage changes.
! These are discussed below.
- a. Higher Core melt temperature. A case was run in which core melt 0
I temperature was assumed to be 3000 K as contrasted to the base case 2500 K. This was intended to delay the onset of core geometry degradation, which in turn provides more time to heat other portions of the RCS. The 5000K change in relt temperature was found to cause 29
only a few degrees change in SG tube temperatures, which was attrib-uted to the extremely rapid terrperature increase rate in the core as melt temperature is approached, and a concomitant small increase in the time available for heat transport to the steam generators.
The model is based upon assumed symmetric behavior, whereas some asymetries have been found experimentally. If these contributed to a preferential flow of hot fluid near one of the hot legs, that leg might transport hot fluid toward a stearu generator and provide higher temperatures than detemined in the calculation. This could increase the computed impact of the sensitivity calculation.
A second aspect of the modeling that would act to reduce the calcu-lated impact of the sensitivity run is the assumption of mixing within the steam generator inlet plenum. We believe an assessment of this effect is needed, as previously identified,
- b. RCP seal failure. RCP seal failure, if it were to occur, was felt to be a leak in the range of 50 gpm (water) per seal. This was modeled, with the break occurring in all four RCPs at 45 minutes after initiation of the accident. This was found to have an insig-nificant impact on the results (Refs. 1? and 18).
PSNH also addressed preexisting leaks in SG tubes which are within technical specifications. These were stated to be small in compari-son to the 50 gpm flow rate associated with seal leaks, and conse-quently were argued as being negligible (Ref.17).
We believe the preexisting leak situation has a negligible impact on NSSS behavior as long as the leak remains small, but do not accept the argument advanced by PSNH as the reason. A comparison of the velocity associated with flow in a tube due to natural circulation with that associated with the leak, with establishing that the latter was negligible, would be more convincing. Similarly, a comparison of 30 l
l
flow rate induced by the RCP seal-rupture to that expected for natural convection flow would be helpful. Further, one would have to l establish that such a leak, passing steam, would not result in steam -
passing through the crossover leg seal at such a rate as to perturb the conclusions.
Provision of temperature-information pertinent to fluid passing
, through the RCP seals would be helpful.
- c. SG secondary side blowdown. Plys (Ref. 18) reports a calculation to investigate the effect of reduced cooling on the SG secondary side in which the steam generator PORVs are assumed to stick open, thus depleting the secondary side of a high pressure steam atmosphere.
Drastic differences were discovered early in the accident due to coolino as the steam generators blew down. Sufficient cooling was provided that the pressurizer emptied due to primary fluid contraction. Reactor vessel failure occurred slightly earlier in this case as contrasted to the base case due to less heat removal from the primary system following removal of'the secondary side heat sink. An initial peak in SG inlet plenum temperature of 860UK is-identical to that of the base case, but occurs about 500 sec earlier.
Following the initial peak, the plenum temperature behavior is similar to the base case, although displaced in time, but is 50 to 1000 K higher over the remainder of the transient.
We suggest the calculation be conducted by assuming the PORV is stuck open after all water has been vaporized. This avoids the situation of overcooling associated with the early opening, and may be more
, compatible with some postulated operator actions associated with late attempts to deal with approaching core melt.
Again, we are concerned with the influence of assumed mixing in the steam generator inlet plenum and the impact upon calculated results.
31 t
u- - --
- r. --,-.e-. -,,--,n,..- , - - - - - - - - , ,
Core resistance variation. Variation of the resistance of the core d.
to flow was evaluated by lowering the axial and cross flow core friction factors in one calculation. This slightly increased heat transfer to the steam generators and correspondingly increased time to vessel failure. There was a slight tube te:nperature increase, but in general, the calculation showed little sensitivity of tube temperature to the change in core friction factors.
- e. Reduced SG tube circulation. Selection of lower limit values of the number of steam generator tubes participating in flow from the inlet to the outlet plena was used for another sensitivity calculation.
This provided lower values of steam generator natural circulation flow relative to the hot leg natural circulation flow rate, and reduced cooling of the steam generator inlet plenum due to flow from the outlet plenum. Slightly less heat was removed from the reactor vessel due -to the lowered flow rates, and vessel failure occurred slightly earlier. These changes were insignificant. However, the steam generator inlet plenum was found to be about 1500K higher than for the base case, reaching a temperature of 9800 K for a short time.
Steam generator tube temperature was relatively unaffected.
Comparison of inlet plenum and tube temperature transient behavior (References 17 and 18's Figures 4-11 and 4-12) appears to indicate a significant thennal inertial associated with the tubes, which do not increase in temperature to a sianificant degree in contrast to the temperature of the source fluid in the steam generator inlet plenum.
We believe this needs further discussion. For example, what is the location of the tube temperature and does this location correspond to the highest tube temperature?
Again, as previously stated, the influence of the assumpt on of complete mixing in the steam generator inlet plenum will impact the results. A portion of the concern is that reduced flow rates may 32
h lead to greeter stratification and less mixing in the SG plena, a phenomenon that is not modeled in the pSNH reported evaluations, and a phenomenon with the potential to increase tube temperatures over what was reported.
- f. Core blockage. In this calculation, a delay of blockage in the core at the time of core melt to the time the node was completely filled with refrozen eutectic was assumed. This was done to continue core oxidation and core / upper plenum flow for a longer time. For this case, the maximum sustained SG inlet plenum temperature is roughly 1060U K, with a short time (less than 50 seconds) temperature " spike" to about 11200K.
We again reiterate the concern with SG inlet plenum modeling and its impact upon the results.
- g. Sensitivity Summary. An approximate comparison of the results of the sensitivity study is provided in Figure 1. The ma.ior early effect on increased tube temperature is due to changing the SG tube flow characteristics. Later, and with the greatest impact, is the effect of delaying formation of blockage in the core, which allows continued circulation of hot fluid through the core where the temperature is increased, as opposed to a drastic reduction in heat transport between the core and other RCS components when a core geometry chance occurs.
4 Steam Generator Tube Strength. Plys, in Reference 18, Appendix B, addresses SG tube integrity. The presentation is based upon the SG secondary side pressure being at the SG safety or relief valve setpoints which, as previously discussed, may not be the case. We note that Plys identifies nominal hoop stresses of 9300 to 10000 psi for the assumed conditions. Hence, the case of the SG secondary being depressurized will 33
1 result in a nominal hoop stress of roughly 19,000 psi. This stress, l
substituted into Reference 18's Figure B-6, results in a Larson Miller l parameter of about 37. The Larson Miller parameter is defined as:
LM = T(20+1og t ) x 10-3 l r
4 where:'
T = temperature, OR t = time to rupture, hrs.
r Substituting a temperature of 10900K (the value used by Plys to conclude the rupture time would be greater than 2.5 hrs) yields a time to rupture of about 5 minutes, a significant change from the Plys value.
Plys could have selected 10900 K as conservative, with no need to consider an alternate since the no tube rupture position was supported by the result. If we recognize this possibility, and select a less conservative 0
1000 K, we find a rupture time of about 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. These temperatures can be contrasted to the SG inlet plenum temperatures provided in Fioure 1, with recognition that these are not tube temperatures, but also with the recognition that some of the parameters contributing to the temperatures remain to be evaluated.
Clearly, we are in a temperature region where relatively small changes have a significant impact upon creep rupture time. Equally clearly, tube
{- stress could be roughly a factor of two higher than the value used to justify that tubes would not rupture. We conclude the picture is not as clear as presented in Reference 18, which presented a conclusion that tubes would not be ruptured.
3.2.8. Other Considerations In Reference 17, PSNH stated that if one postulated creep rupture failure of steam gererator tubes, the pressure inside the previously dried out and isolated steam generator secondary side would increase until the steam 34 1
.. ~ _ - -
4 generator PORV's setpoint was reached, at which time the valves would lift and modulate until reactor vessel helt through and RCS depressurization into the containment. During the periods of SG PORV opening, there would be a high leak rate bypass condition directly from the RCS to outside the i
containment. They further stated that after vessel melt through, the-leak rate out this _ path would be low and would correspond to any low pressure leakage through the reclosed PORV. They note this leak path could-be
- enhanced if the SG safety valves also lift and fail to reseat properly; however, they believe it unlikely that the safety valve setpoint would be reached.
As previously discussed, we do not believe an individual tube would rupture, but instead believe there would be a massive failure in one steam generator. (Once the failure initiated, we would expect the RCS to depressurize rapidly, which would reduce stress on tubes in other steam generators.) It is difficult to postulate a PORV modulating this condi-tion. It is further difficult to postulate the PORV or the safety valves would not be damaged when exposed to these conditions, and therefore their reclosing may be questionable. One may also question SG secondary side structural integrity when exposed to the high temperature environment.
- Finally, if the conditions which led to the accident sequence involve a l
loss of all AC power, which is one of the likely situations given a severe accident scenario, we pose the question of how lona the PORVs can be
- expected to modulate pressure assuming they are not damaged by the fluid l being modulated.
( Plys (Ref. 18) has identified that the MAAP code does not model certain aspects of SG tube temperature, and a method of obtaining temperature was discussed. Aside from the impact of secondary side steam as a cooling
, medium, we are concerned about local heating due to small leaks. Such a leak could cause a small amount of hot fluid to pass through a localized area into the SG secondary side, with different heat transfer character-istics and tube temperatures than one would encounter with the treatment of overall inside to outside heat flow utilized by Plys in their estima-tion. Whether this is important to localized tube temperature over a 35 i
,._ ~ .__
sufficient area to be of concern should be addressed. (Note the effect could also be concentrated in an adjoining tube. This can be visualized by picturing a tube with a small hole which directs hot RCS fluid onto the secondary side surface of an adjoining tube, while the inside surface of that same tube is exposed to hot RCS fluid.)
3.3 Accident Likelihood PSNH has estimated the mean annual frequency of accidents in which the core melts with the RCS at high pressure and the SGs dry as bounded by a value of 4.5 X 10-5 per reactor year (Ref.17). This is composed of the following plant damage states:
Plant Damage Mean Annual State (PDS) Frequency 3D 1.5 X 10-5 3FP 8.9 X 10-6 4A 1.4 X 10-5 4C 1.7 X 10~7 4D ?.8 X 10-611 4E 2.2 X 10 7 4FP 1.2 X 10-8A 3.9 X 10-6 Total 4.5X 10 -5 The accident sequences which comprise the PDSs include transient and loss of off site power sequences with failure of all emergency feedwater, failure of feed and bleed with loss of all emergency feedwater, and transients without scram. PDS 8A consists of eight sequences which involve station blackout and emergency feed water failure with recovery of containment heat removal.
PSNH also addresses the potential impact of tube rupture on this information.
They have assigned a hiah chance of no containment failure to PDSs A. PDSs C and D are considered as leading to a high likelihood of long term containment overpressure failure. PDSs FP are a hiah chance of small bypass, and PDS E is 36
a high chance of large bypass. Hence, PDSs A, C, and D would be impacted by SGTR, and FP may represent some impact. Addition of the appropriate values indicates that the likelihood of being in a condition where SGTR could affect the results is about 4 X 10-5 (as contrasted to the assumption of no SGTR).
PSNH considers these values to be bounding because some of the values inc16 states with water on the steam generator secondary side, for which SGTR is not a concern, certain operator recovery actions have been neglected, and RCS depressurizations prior to core melt have not been considered. As previously discussed, operator depressurization is one of the potential steps which one could consider to mitigate SGTR. PSNH estimates the frequence of operator failure to depressurize as.less than 10-2 to 10-3 per demand, provided proce-dures are modified and adequate operator training is provided. Using these '
values results in a frequency of obtaining conditions under which SGTR would be of concern of about 10-7 to 10-8 per reactor year.
- Although these values appear reasonable, we note that the conditions which led to the factor of 10~2 to 10-3 reduction do not presently exist. We further would need substantiation for these values prior to acceptance.
j Discussion is also provided concerning the likelihood of SGTR if exposed to high pressure core melt conditions (Ref.17). PSNH points out that their calculations show SG tube temperatures that are roughly 200 to 300 F belor what would be required for creep rupture, and this is identified as principally due to cooling by steam on the SG secondary side. Several things are necessary for acceptance of the tube temperature conclusions, including, as discussed elsewhere, substantiation of the calculational technique and investigation of the likelihood of the SG secondary side having a significant steam inventory (which also means having a significant pressure).
Finally, PSNH estimates a 99% chance thct failure of SG tubes will not occur before reactor vessel melt through or piping noz;:le failure. This value, combined with the prior PSNH estimates of frequencies, appears sufficient to i
37 l
l t
. .. .- - . . _ _ _ _ . _ . _ , _ - _ _ . _ ~ . . . _ _ _ _ _ _ _ . . _ _ _ . - . . _ _ _ . _ , . . . . _ . _ , _ _ _ . _ _ - - - . _ -_ _ - -_
establish that SGTR is not of concern as a significant contributor to risk.
Therefore, one can reasonably anticipate that substantiation of the various items which led to the conclusion, as discussed in this comunication, should provide substantiation of the above preliminary conclusion.
3.4 Additional Observations A number of observations and conenents have been made in the previous discus-sion, k'e offer the following additional coments:
L Much of the modeling utilized in the calculations has not been documented.
We understand this is underway. Such documentation will be helpful in the dontinuation of the review.
- 2. Thi outside of the hot legs is assumed to be adiabatic. This probably introduces a small conservatism into the results with respect to hot leg l temperature. The impact on other parameters is probably negligible. With respect to the hot legs, the parameter of interest may involve a relatively thin wall connecting pipe that is exposed to high fluid temperature, and whose temperature will follow fluid temperature more closely than is the case with the relatively massive hot leg: or the vessel nozzle region of the hot leg, which will be more closely allied with fluid circulating rapidly within the upper plenum. Thermal response of these regions may be critical in determination of the failure point of the RCS pressure boundary.
- 3. Although the limited experimental evidence reveals some symmetry in flow behavior within the reactor vessel, there are also unsymmetrical flows and temperatures. We understand the MAAP calculations are based upon modeling the upper plenum fluid as a single volume. This appears to be a nonconservative approach.
38
4 STEAM GENERATOR TUBE RUPTURE CONCLUSIONS The above discusseo considerations lead us to the conclusion that this topic is in a developing state, with knowledge being rapidly accumulated. Insufficient infomation is presently available for one to conclude that SGTR cannot occur as a result of severe accident conditions.
Our judgement, at this juncture, is that a carefully conducted and thorough evaluation on the part of PSNH, that utilizes information which either exists or will be available within the near future, can establish that the likelihood is small that a SGTR will result due to overheating during severe accidents.
Further, our judgement is that the risk associated with SGTR can be shown to be negligible for these conditions. Our judgement needs to be substantiated. We have encountered too many unanswered questions, unsubstantiated assumptions, 3
and potential conditions which could lead to calculation of increased N temperature to accept a conclusion that SGTR will not occur under circumstances such that the associated risk can be neglected. We further judge that coverage of all areas subject to question will be a substantial task. We note, as a qualifier to these conclusions, that our review is not complete, and, in addition, work is ongoing to provide further informatior Existing knowledge would support a conclusion that SGTR is not a problem if the RCS is depressurized. Consequently, reasonable assurance that progressions toward core melt would not occur at high RCS pressure, coupled with suitable technical backup for a conclusion that low pressure is not of concern, would eliminate our concern regarding SGTR under severe accident conditions. We have not conducted an evaluation of the trade-offs associated with such an approach, nor have we been provided with information that would either support or negate RCS depressurization under severe accident conditions. We have not provided a recommendation regarding whether RCS depressurization is attractive when all pertinent factors are considered due to lack of a balanced picture.
Determination of the correctness of a judoement that SGTR is not a concern under severe accident conditions with the RCS at hiah pressure can be based upon a combination of analytic and experimental investigations. The ongoing test at Westinghouse in which reasonably close similitude is claimed between 39 i
the test facility and appropriate parts of a Westinghouse four loop NSSS should provide key data which can be applied to assist in the confirm. tion of analysis techniques. Selected test data from other facilities and further examination of the analysis technioues, coupled with necessary changes when they are uncovered, should provide sufficient confirmation that reasonable reliance can be placed upon accident analyses pertinent to this issue. Application of a reliable analysis technique to issue investigation should then provide the necessary background to resolve this. issue. Such a program will represent a formidable undertaking.
e 40
9
- 5. REFERENCES
- 1. "Seabrook Station Probabilistic Safety Assessment," Pickard, Lowe and Garrick, Inc., PLG-0300, December 1983.
- 2. Garrick, John B., Karl N. Fleming, and Alfred Torri, "Seabrook Station Probabilistic Safety Assessment, Technical Summary Report," Pickard, Lowe and Garrick, Inc. , PLG-0365, June 1984.
- 3. "Seabrook Station Risk Management and Emergency Planning Study", Pickard, Lowe and Garrick, Inc., PLG-0432, December 1985.
- 4. "Seabrook Station Emergency Planning Sensitivity Study", Pickard, Lowe and Garrick, Inc., PLG-0465, April 1986.
- 5. " Reactor Safety Study: An Assessment of Accident Risks in U. S. Comercial .
Nuclear Power Plants," U. S. Nuclear Regulatory Commission, WASH-1400, NUREG-75/014, October 1975.
- 6. Winters, L., "RELAP5 Station Blackout Transient Analysis in a PWR," ENC Memo No. 8.904.00-GR17, July 1982.
- 7. Denny, V. E., and B. R. Sehgal, " Analytical Prediction of Core Heatup/Liquefication/ Slumping," Paper TS-5.4, Proceedings Intl. Meeting on LWR Severe Accident Evaluation, Cambridge, MA, August 28 - September 1, 1983.-
- 8. Lyon, Warren C., " Report on Meeting to Discuss RCS Pressure Boundary Heating During Severe Accidents (May 14,1984)," NRC Memorandum to Distribution, June 15, 1084.
- 9. Bernero, Robert M., "Need for Multidimensional Modeling of RCS Behavior in l Support of Severe Accident Investication," NRC Memorandum for Denwood F.
Ross, August 30, 1984.
- 10. Sheron, Brian W., " Steam Generator Tube Response Durino Severe Accidents,"
NRC Memorandum to B. D. Liaw, February 14, 1985.
l 11. Stewart, W. A., A. T. Pieczynski, and V. Srinivas, " Experiments on Natural l Circulation Flow in a Scale Model PWR Reactor System during Postulated Degraded Core Accidents," Paper 10.C, Proceedings of Third International Topical Meeting on Reactor Thermal Hydraulics, Newport, RI, October 15 -
18, 1985.
i
} 12. P1ys, Martin G., Marc A Kenton, Robert E. Henry, and Peter Kirby, "Seabrook Steam Generator Integrity Analysis," Information presented at Brookhaven National Laboratory by Fauske & Associates, Inc. and Westing-house Electric Corporation, October 17, 1986.
- 13. Stewart, W. A., A. T. Pieczynski, and V. Srinivas, " Experiments on Natural Circulation Flow in a Scale Model PWR Reactor System During Postulated Degraded Core Accidents," Westinghouse Electric Corporation, Pittsburgh, PA, Scientific Paper 85-5J0-RCIRC-P2, August ?9,1985.
41
- 14. Stewart, W. A., A. T. Pieczynski, and V. Srinivas, " Experiments on Natural Circulation Flows in Steam Generators During Severe Accidents,"
Westinghouse Electric Corporation, Pittsburgh, PA, Scientific Paper 85-5J0-RCIRC-P3, December 5, 1985.
- 15. Chen, B. C-J. ,11. M. Domanus, W. T. Sha, and B. R. Sehgal, " Degraded Core Study Using the Multidimensional COMMIX Code," Trans. ANS. Vol. 49, pp.
453-454, June 1985.
- 16.
Dearing,
J. F., " Flow-Pattern Results for a TMLB' Accident Sequence in the Surry Plant Using HELPROG," Los Alamos National Laboratory, LA-UR-85-3668, November 1985.
- 17. DeVincentis, John, " Response to Request for Additional Information (RAIs)," Letter from Public Service of New Hampshire to Steven M. Long of NRC, SBN-1227, T.F. B7.1.2, November 7, 1986.
- 18. Plys, Martin G. , et. al. , "Seabrook Steam Generator Intearity Analysis,"
Fauske & Associates, and Westinabouse Electric Corporation, November, 1986. (Provided via Reference 17.)
- 19. Theofanous, T. G., " Severe Accident Containment Phenomenology for Probabilistic Risk Assessment", handout pertainino to a presentation to an ACRS Seminar, March 22, 1984.
2D. Theofanous, T. G., and Chien-Hsuing Lee, "The Direct Heating Problem",
presentation to the Containment Loads Working Group Meeting, Rockville, Md.,
March 1984.
- 21. Nourbakhsh, H. P., et al. " Natural Circulation Phenomena and Primary System Failure in Station Blackout Accidents", Section 24 of " Proceedings:
The Sixth Information Exchange Meeting on Pebris Coolability", Meeting held Nov. 7-9, 1984, EPRI NP-4455, March 1986.
- 22. Theofanous, T. G., " Review Coments on Seabrook Station Steam Generator Tube Response During Severe Accidents (a draft NUREG Report dated 12/15/86) and related sections of Technical Evaluation of the EPZ Sensitivity Study for Seebrook (a draft BNL report dated 12/5/86)", Dept. of Chem. and Nuc. Eng., Univ. of Calif., Santa Barbara, CA 93106, Jan. 12, 1987 42 L
o --.
TECHNICAL EVALUATION OF THE EPZ SENSITIVITY STUDY FOR SEABROOK W. T. Pratt and C. Hofmayer Principal Investigators Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 March 1987 l l
s Prepared for U.S. Nuclear Regulatory Commission Washington, DC 20555 Under Contract No. DE-AC02-76CH00016 FIN A-3852
1
- L w LIST OF CONTRIBUTDRS Contributor Affiliation K. Bandyopadhyay Structures & Components Evaluation Group / SAD P. Bezler Civil & Structural Mechanics Group / SAD I G. Bozoki Risk Evaluation Group /SRED ,
T-L. Chu Risk Evaluation Group /SRED. !
M. Chun Accident Analysis Group /SRED I C. Hofmayer Structures & Components Evaluation Group / SAD M. Khatib-Rahbar Accident Analysis Group /SRED
- 8. Luckas Engineering Analysis & Human Factors Group /ETD J. Pires Structures & Components Evaluation Group / SAD W. T. Pratt. SRED A. Tingle Accident Analysis Group /SRED R. Youngblood Facilities Risk Analysis Group P. C.. Wang Civil & Structural Mechanics Group / SAD SRED = Safety and Risk Evaluation Division ETD = Engineering Technology Division SAD = Structural Analysis Division 9
iii
ABSTRACT A technical evaluation of the Seabrook Station Emergency Planning Sensi-tivity Study (PLG-0465) and supporting documentation has been performed. This was an evaluation which focused on those areas found to' be the most influen-tial in calculating the Seabrook risk estimates. The approach taken by Brook-haven National Laboratory (BNL) was to perform sensitivity studies to assess the impact on the results in PLG-0465 of the BNL evaluation of these areas.
e a
i V
l TABLE OF CONTENTS Page List of Contributors.................................................... iii Abstract................................................................ V Li s t o f Fi gu r e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i x L i s t o f Ta b l e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x i i Preface................................................................. xiii Acknowledgements........................................................ xv S u mma ry . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x v i i
- 1. INTRODUCTION....................................................... 1-1 1.1 Background.................................................... 1-1 1.2 . Scope and Focus of Revi ew. . . . . . . . . . . . . . . . . . . . . . . . . . . .'. . . . . . . . ' . 1-4 .
1.3 Ap p r o a c h t o Re v i ew . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.4 Organization of the Report.................................... 1-7 1.5 References.................................................... 1-8
~1
- 2. SYSTEM EVALUATION.................................................. 2-1 2.1 In t e r f ac i n g Sy s t em L0C A. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 ;
2.1.1 G e n e r a l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -2 l 2.1.2 Ot h e r I S L P a t h s . . . . . . . . . . . . . . . . . . . . . . . . '. . . . . . . . . . . . . . . . 2 - 3 i 2.1.3 ISL Ini ti ator Frequenci es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 2.1.3.1 Check valve failure frequencies;.............. 2-5 2.1.3.2 Cold leg safety injection path frequency...... 2-8 2.1.3.3 RHR suction side frequency.....................-2-9' 2.1. 4 . . Op e ra t o r Acti on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-11 2.1.5 Break Location.......................................... 2-15 2.1.6 Event Tree Quanti fi cati on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-16 2.2 . Accidents During Shutdown and Refueling Conditions............ 2-17 2.2.1 Loss of Decay Heat smoval During Shutdown or Refueling. 2-20 2.2.2 Low Temperature 0verp ressuri zation. . . . . . . . . . . . . . . . . . . . . 2-23 2.2.3 Loss of Coolant iccidents During Shutdown or Refueling., 2-24 2.2.4 PSNH. Comments on BNL Draft Report...................... 2-25 2.2.5 Summary of the Shutdown Ri sk Revi ew. . . . . . . . . . . . . . . . . . . . 2-26 2.3 Induced Steam Generator Tube Rupture (SGTR) . . . . . . . . . . . . . . . . . . . 2-27 2.4 Cont a inment Is ol at i on Fa i l ure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-28 2.5 Summary....................................................... 2-29 2.6 References.................................................... 2-33
- 3. EVALUATION OF CONTAINMENT BEHAVIOR................................. 3-1 3.1 Capaci ty a t Gene ral Yi el d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2 Behavi or at large Deformati on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 3.3 Capabi l i ty of Penetrati ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-9 3.4 Summa ry of St ructural Fi nd i ngs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-15 3.5 References.................................................... 3-16 vii
- b. .
- TABLE OF CONTENTS (Continued) 1
b Page 4 CONTAINMENT EVENT TREE REVIEW...................................... 4-1 4.1 Potenti al Co nt ai nment Load s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.2 Appl i cati on to Seab rook . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.3 Su mma ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 5 4.4 ' References.................................................... 4-6
- 5. REV I EW OF SOURCE TERMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 5.1 Fi del i ty to WASH-1400 Methodol ogy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Credit for Scrubbing of Submerged Rel eases . . . . . . . . . . . . . . . . . . . . 5-1
- 5. 3 Un c e rt a i nt i es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -2 5.4 Su mm a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 3
- 5. 5 Re fe ren ces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -3
- 6. S ITE CONS EQUE NCE N0 DEL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 NUR E G - 0 396 8a s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 -1 6.2 Cons eq uence Model i n g. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 l 6.2.1. Whole Body Dose Vs Di stance. . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.2.2 Thyroi d Dose Vs Di stance. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.2.3 Ri sk of Ea rly Fatal iti es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 6.3 Comp a ri sons of Resul t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5
. 6.3.1 Resul ts of Seabrook Study. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5
, 6.4 Sens i ti vi ty St ud i es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6.4.1 Sensitivity.of Results to Multipuff Release............
6-7 6 . 4 . 2 , Su mma ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 7 6.4.2.1. Interfacing systems L0CA...................... 6-7
, 6.4.2.2 Acci dents du ri ng shutdown. . . . . . . . . . . . . . . . . . . . . 6-8 6.4.2.3 Induced steam generator tube rupture.......... 6-10 6.4.2.4 Containment isolation. failure and pre-exi sti ng l eakage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 6.4.2.5 Contai nment structural capacity. . . . . . . . . . . . . . . 6-12 6.4.2.6 Containment loads............................. 6-13 6.4.2.7 So u r c e t e rms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 - 15 6.4.2.8 Consequence Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-16 6.5 References.................................................... 6-17 viii
. o 4-l LIST OF FIGURES .
Figure Page i S.1 Comparison of Secbrook Station sensitivity results using WASH-1400 source term methodology with background, safety goal i ndi vi du al 'and RMEPS ri sk 1 evel s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxx S.2 Comparison of median risk of early fatalities at Seabrook .
Station for. different emergency planning options.. .. ... ..... . . .. xxxi l S.3 Comparison of Seabrook Station results'in this study and RMEPS ^
- with NUREG-0396 - ~200-rem and 50-ren whole body dose plots for no i mmedi ate protecti ve acti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxxi i S.4 200. rem dose versus distance curves for various failure modes -
assumi ng no _ immediate protecti ve action. . . . . . . . . . . . . . . . . . . . . . . . . xxxii i S.5 Comparison'of 200 rem-dose ~versus distance curves for conservative assumption of.no credit for operator recovery of
, open equ i pment hatch . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xxxi v
-S.6.
Comparison of.8NUREG-0396.....,NL sensitivity' studies with PLG-0465 and
................................................. xxxv }
S.7 Comparison of 200 rem-dose versus -distance curves for '
conservative interpretation by PSNH of NUREG/CR-4220 data....... ,xxxvi l 1.1 Comparison of Seabrook Station sensitivity results using WASH-1400 source term methodology with background, safety goal
, i ndi vidual a nd RMEPS ri sk 1 evel s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-9 ,
4 1.2 Comparison of median risk of early fatalities at Seabrook 4
Station for di fferent emergency planning options.. .. . .. ... ..... 1-10 '
1.3 Comparison of Seabrook Station results in this study and RMEPS with NUREG-0396 - 200-rem and 50-rem whole body dose plots for no innedi ate p rotecti ve acti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-11 4
1.4 200-rem dose versus distance curves for various failure modes assuming no immediate protective action........................ 1-12 2.1 Frequency of accumulator check valve leakage events............ 2-36 2.2 Comparison of 200 rem-dose versus distance curves for
{ conservative assumption of no credit for operator recovery
, of open equipment hatch........................................ 2-37 2.3 Comparison of BNL sensitivity studies with PLG-0465 and NUREG-0396..................................................... 2-38 2.4 Comparison of 200 rem-dose distance curves for conservative i nterpretati on by PSNH of NUREG/CR-4220 data. . . . . . . . . . . . . . . . . . . 2-39
- 3.1 Containment building cross-section............................. 3-17 3.2 Cylinder reinforcement......................................... 3-18 3.3 Containment finite element model (NFAP)........................ 3-19 3.4 Pressure-radial displacement relation for containment.......... 3-22 6.1 Components of NUREG-0396 curve as computed by BNL using CRAC2.. 6-19 i 6.2 Risk of death or exceeding dose levels for S1W as calculated by BNL......................................................... 6-20 6.3 Risk of death or exceeding dose levels for S2W as calculated by BNL......................................................... 6-21 6.4 Risk of death or exceeding dose levels for S6W as calculate by BNL......................................................... 6-22 ix
o .
LIST OF FIGURES (Continued)
Figure Page 6.5 Dose versus distance curve for release category S1W from Seabrook for no immediate protective action with BNL results usi ng MACCS superimpos ed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-23 6.6 Dose versus distance curve for release category S2W from Seabrook for no immediate protective action with BNL results usi ng MACCS superi mpos ed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-24 6.7 Dose versus distance curve for release category S6W from Seabrook for no immediate protective action with BNL results using MACCS superimposed....................................... 6-25 6.8 Compari s on of MACCS to CRAC2 codes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-26 6.9 Comparison of 200 rem-dose versus distance curves for conservative assumption of no credit for operator recovery o f open equi pme nt hat ch . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 7 6.10 Comparison of BNL sensitivity studies with PLG-0465 and NUREG-0396........................................................... 6-28 6.11 Comparison of 200 rem-dose versus distance curves for conservative interpretation by PNSH of NUREG/CR-4220 data...... 6-29 s
X 4
. LIST OF TABLES-9 Table Page S.1 -Summary of Release Category Frequency Uncertainty Distributions. xxxvif S.2 Early Fatalities Conditional on a Release Occurring in the Population Around the Seabrook Station Site Boundary........... -xxxviii-1.1 Summary of Release Category Frequency Uncertainty Distributions.1-13 1.2 - Risk of Early Fatalities in the Population Around the Seabrook St a t i on S i t e Bou nd a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -14 2.1 Summary of Operating Events, Emergency Core Cooling System, Isolation Check Valves, Leakage Failure Mode...................
2-40 2.2 _ Summary of Operating Events, Emergency Core Cooling System, Isolation Check Valves, " Failure to Close U Demand" Failure Mode...............................pon .................... 2-43 2.3 Accumulator Check Valve Exposure Data.......................... 2-44 2.4 . Statistical Data on Leakage Events of Check Valves to AC C u mu l a t o r s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 4 5 2.5 ISL Results Initially Assigned Plant Damage States............. 2-46 2.6 P l a nt Op e ra t i n g Mod es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -4 7 2.7 Categories of 130 Reported Total DHR System Failures When Required to Operate (Loss of Function) at U.S. PWRs: 1976-1983.. 2-48 3.1 Statistics of Rebar Yield Strength for Various Sizes........... 3-21 3.2 Reinforcement Details of the Containment Cylinder.............. 3-22 3.3 Rein forcement Det ai ls of the Contai nment Dome. . . . . . . . . . . . . . . . . . 3-23 3.4 Statistics of Concrete Compressive Stren 3-24 4
3.5 Concrete Properties.....................gth.................... ....................... 3-25 3.6 Characterization of Containment Penetrations................... 3-26 4.1 Comparison of Core Melt Frequencies and Distribution of Release Types..........................................................4-7
' 5.1 Release Categories for Seabrook Station Based on WASH-1400 Sou rce Te rm Met hod ol ogy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 5.2 Revi sed C-Mat ri x for New Sou rce Te rm Categori es . . . . . . . . . . . . . . . . 5-6 6.1 Summary of Release Categories Representing Hypothetical Accidents (from the RSS)....................................... 6-30 l
f l
I i
l xi
PREFACE This report describes a technical evaluation of the Seabrook Station Emergency Planning Sensitivity Study (PLG-0465) and the Seabrook Station Risk Management and Emergency Planning Study (RMEPS) (PLG-0432). The main objec-tive of this technical evaluation is to assist the NRC in its evaluation of the validity of the conclusions presented in PLG-0465. This is therefore a '
focused review by Brookhaven National Laboratory (BNL) of those areas identi-fied in PLG-0465 as being the most influential in calculating the Seabrook risk estimates. However, regardless of the conclusions of this focused review, BNL cannot attest to the validity of the overall risk profiles pre-sented in PLG-0465. This follows from the observation that the risk estimates in PLG-0465 rely heavily on RMEPS, which in turn relies on the Seabrook Sta-tion Probabilistic Safety Assessment (SSPSA). Unfortunately, the risk pro-files in the SSPSA have not been independently reassessed, requantified, and validated, by the NRC staff or their contractors. Similarly, within the scope of the review, BNL has also not validated the accident sequence probability
- estimates in the SSPSA. Therefore, because these estimates form the founda-tion for the updated risk estimates in the RMEPS and ultimately in PLG-0465, BNL has not verified the total risk estimates in PLG-0465. This includes the predicted dose versus distance curves. The current review should therefore be regarded as an evaluation of selected issues related to the potential for a large early release of radioactivity at the Seabrook Station. It is not a reassessment or validation of the total risk profile.
5 xiii
L e r ACKNOWLEDGEMENTS The authors wish to thank Dr. R. Bari and Dr. M. Reich in the Department of Nuclear Energy at Brookhaven National Laboratory for many discussions, com-ments, and suggestions related to this program.
This work was performed for the Division of PWR Licensing - A, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission. The authors wish to acknowledge G. Bagchi, S. Long, W. Lyon, and S. Newberry for their support and guidance throughout the course of this program.
Lastly, the -authors acknowledge the efforts of C. Conrad, A. Costini, E. Gilbert, and D. Votruba in preparing this document for publication.
XV
l
SUMMARY
~
7 4 .
- 11. ' Purpose
'This report describes 'a technical' evaluation by Brookhaven National
. " Laboratory (BNL) of the Seabrook Station Emergency Planning: Sensitivity Study (PLG-0465), which was prepared for the Public Service Company of New Hampshire
'(PSNH) by Pickard, Lowe and =Garrick:(PLG) Inc. PLG-0465 reviewed the bases in NUREG-0396 for the current 110 mile emergency planning zone' (EPZ) and argued, by :taking into . account Seabrook-specific ' plant features and improvements in methodology - that a . smaller EPZ- was justified .at Seabrook. The results in PLG-0465 rely heavily on two earlier studies, namely the Seabrook Station Risk Management and Emergency Planning Study (RMEPS)- (PLG-0432) and the Seabrook Station Probabilistic Safety Assessment (SSPSA). PLG-0432 and RMEPS were also prepared by PLG for PSNH.
The SSPSA was an extensive evaluation of the risk associated with opera-tion of the Seabrook Nuclear Power Station. The SSPSA investigated the conse-quences of ' accidents ' that might occur from initiating events that could: be internal .to the plant 'and also external (e.g., seismic events). The Nuclear Regulatory . Commission' (NRC) and its supporting contractors initiated an
'in-depth review of the SSPSA but this effort was terminated prior to requan-tifying the risk estimates. '
- RMEPS' was a sensitivity study which evaluated emergency planning. options j for the Seabrook Station. RMEPS focused on those areas that were found in the f SSPSA- to be the leading contributors to risk. RMEPS rebaselined the SSPSA i analysis of these risk-important areas specifically .to establish an updated assessment of the risk - at TSeabrook so that alternative emergency planning options could be developed. Therefore, RMEPS focused on new data and t engineering insights about the initiation and progression of sequences involv-ing interfacing systems loss .of-coolant accidents (LOCAs) and on- the results
- of experimental and . analytical research that provide an enhanced basis for i
assessment of radioactive material release (i.e.,' source terms) for a wider i
spectrum of accident sequences. Thus, RMEPS _ represents the applicant's best i estimate.of risk at Seabrook. RMEPS-is being reviewed in conjunction with the review of.PLG-0465.
i The sensitivity studies in PLG-0465 were performed to determine the ,
extent to which the conclusions of RMEPS were dependent on any new source ter:n ,
technology. Thus, PLG-0465 only changed the source terms in RMEPS to be
! consistent with WASH-1400 source term methodology. Therefore, in order for ;
BNL to evaluate the results and conclusions in PLG-0465 we also had to evalu-ate the results in the RMEPS and the SSPSA. It will be shown later that the l current BNL evaluation focused on the results in PLG-0465 and the RMEPS and did not attempt to review in detail the results of the SSPSA. This in turn led to the need to qualify the conclusions of the BNL review of PLG-0465.
! The principal conclusion of PLG-0465 was that an EPZ at the Seabrook Sta-
! tion of 1 mile radius or less is more justified in terms nf its risk manage-
! ment effectiveness than the current 10-mile EPZ was justified by the results i of NUR EG-0396. This conclusion was based on the results of the PLG-0465 l Sensitivity _ Study, which are reproduced in Figures' S.1-S.3. These results l
! xvii l
1 f
were constructed without accounting for any new insights about source terms since WASH-1400. The conclusion was based on the following observations:
The individual risk of early fatalities in the population within 1 mile of the site boundary with no immediate protective actions is less
- than the NRC safety goal (refer to Figure S.1). This individual risk is substantially less when a 1-mile" evacuation is assumed.
The risk of early fatalities with a 1-mile evacuation is comparable to the WASH-1400 results, which assumed a 25-mile evacuation (refer to Figure S.2).
Thethan substantially less Seabrook Station those for results for a 2-mile evacuation are WASH-1400 The risk of radiological exposures for 1, 5, 50, and 200-rem whole body doses with no immediate protective actions is less at 1 mile than the corresponding NUREG-0396 results at 10 miles (refer to Figure S.3). ,
The above observations led to the statement in PLG-0465 that "there is no significant sensitivity frequency results." of exceeding 200 rem beyond 1.5 miles in the Seabrook PLG-0465 identified the following three areas as being {
the most influential in calculating the Seabrook risk estimates:
The effectiveness of the Seabrook Station primary containment to either remain intact or to maintain its fission product retention capability for periods hoc protective actions.much longer than required for even delayed, ad A more realistic assessment of the strength and failure modes of the Seabrook containment than was possible within the state-of-the-art of PRA when the RSS was completed.
A more realistic treatment of the initiation and progression of inter-facing systems LOCA sequences.
Note that of the three areas identified above as being the most influen-tial to the risk estimates in PLG-0465 the firt.c relates to design features that are specific to Seabrook, namely the structural capacity of the Seabrook containment. The other two areas refer to improvements that have been made in our ability to perform risk assessments, which therefore have application to other nuclear power plant risk assessments.
- 2. Rational for Review At the request of the NRC, the BNL technical evaluation initially focused on the following areas in PLG-0465 and RMEPS:
- Interfacing systems LOCAs
- Containment function:
- Isolation failure
- Pre-existing leakage
- Structural capacity
- Containment loads
- Seabrook-specific WASH-1400 source t'erms
- Site consequence model.
xviii
9 However, during the review process additional areas that were outside of the scope of' the original BNL; review were-identified as potentially important to risk at Seabrook. Two of these areas were considered sufficiently impor-tant to Pequest the applicant to provide additional information on the risk associated with such events. The two areas identified were:
. - accidents during shutdown
-potential for induced steam generator tube rupture.
These areas were not initially included in the original BNL review because- in the past they were not found to be dominant risk contributors. -
However, as the risk estimates in PLG-0465 and the RMEPS are relatively low, events that were previously considered to be unimportant now have the poten-tial - to influence the Seabrook risk estimates. Thus, simple sensitivity studies were performed by the applicant and BNL to assess the potential influ-ence of these events on the risk estimates presented in PLG-0465.
By including consideration of these two additional areas (in addition _to the other areas included in the original scope of the BNL review), it should not be assumed that BNL has performed a detailed and systematic search for all events that might be important to risk at Seabrook. Such a search is beyond l the scope of the current BNL review. The current review should therefore be I regarded as an evaluation of selected issues considered important to assessing l the validity of the results and conclusions in PLG-0465 and the RMEPS. The i review therefore focused on assessing ways in which the Seabrook containment may fail or be bypassed early during a severe core melt accident.
3.- Findings in Each Review Area The approach taken by BNL was to perform sensitivity studies in selected areas to assess the impact on the results in PLG-0465 of the BNL review. The BNL sensitivity studies used the conditional risk indices provided in PLG-0465 (and supporting documentation) to assess how changes in the probability of accident sequences 'and containment failure modes would change the Seabrook risk estimates. The sensitivity studies calculated revised 200 rem-dose versus distance curves for comparison with those given in Figure S.3 and revised estimates of individual risk of early fatalities within 1 mile of the site boundary for comparison with the information given in Figure S.I.
The dose versus distance curves in Figure S.3 were constructed from dose versus distance curves (given in Figure S.4) for each of the source terms developed in PLG-0465. These curves were then multiplied by their respective probabilities (given in Table S.1) and summed. The combined dose versus distance curve was then normalized to the total core melt frequency. To be consistent with the NUREG-0396 approach, which used median prooabilities taken from WASH-1400, Figure S.3 was based on the median probabilities given in Table S.I.
The information on individual risk of early fatalities within 1 mile of the Seabrook site boundary given in Figure S.1 is based on the conditional risk indices given in Table S.2 for the various PLG-0465 source terms. The earlier fatality risks were multiplied by the mean frequencies in Table S.1, summed, and then divided by the population at risk. Mean frequencies were i used for this risk measure to be consistent with the NRC safety goal.
XiX l
The BNL review used the information in Table S.2 and Figure S.4 to assess how changes in the probability of accident sequences and failure modes (and hence the probabilities of the source terms in Table S.1) would change the risk estimates given .in Figures S.1 and S.3. Note that Table S.2 also gives the total 'early fatality risk for each release category and that these were the only risk measures available to BNL when the first draft of this report ,
was issued for review. Thus, the preliminary sensitivity studies in the draft report used total early fatality risk and tried to infer how changes in total.
risk might reflect changes in the early fatality risk within 1 mile of the Seabrook site boundary. Hcwever, by comparing the early fatality risk within 1 mile of the site boundary with the total risk it is clear that virtually all of the early fatality risk for release category S2W and-more than half of the risk for release category S6W occurs within 1 mile. It is also clear that most of the risk of early fatalities for release category S1W occurs beyond 1 mile of the site boundary. Therefore, a 2-mile evacuation eliminates all early fatality risk for release category S2W and virtually all for release category S6W. However, a 2-mile evacuation has no impact on the early fatality risk for release category S1W. Thus, it can be misleading to use the total risk of early fatalities as an indicator of the early fatality risk within 1 mile of the site boundary and this led to some confusion in the earlier draft, which has been corrected in this final version of the report.
When mean or median probabilities are used, a range of probabilities is obviously implied and the safety goal specifically states that an attempt has to be made to quantify the uncertainty associated with risk estimates. The applicant considers the WASH-1400 source terms used in PLG-0465 to be very conservative and has a high confidence that the source terms would not be exceeded in a real accident. Therefore, in the opinion of the applicant, only uncertainty in the probabilities of the accident sequences and containment failure modes would impact the risk estimates in Figures S.1-S.3. The appli-cant's upper bound or 95th percentile frequencies, which include consideration of the above uncertainties, are given for each of release category in Table S.I. .
The impact of the 95th percentile frequencies in Table S.1 on the risk estimates in Figures S.1-S.3 is not great. The leading contributor to the risk of early fatalities without evacuation in Figure S.1 is release category S2W. The mean frequency of release category S2W increases by a factor of 5 if the 95th percentile value is used. Therefore, the early fatality risk without evacuation would increase by about a factor of 5 if the 95th percentile fre-quencies were used. However, if 1 mile evacuation is assumed, use of the 95th percentile frequencies would result in an early fatality risk below the safety goal. Also, release category S2W is the only contributor to the 200-rem dose versus distance curves in Figure S.3 and, as this release category has no significant probability of exceeding 200 rem beyond 2 miles (refer to Figure S.4), changing its probability would not significantly change the results in Figure S.3.
In the following sections, the BNL findings related to each review area are briefly summarized. BNL has attempted to follow the ground rules for comparison purposes (namely mean frequencies for comparison with the safety goal and median frequencies for comparison ,with NUREG-0396 results), however, we have also attempted to include a discussion on the uncertainties associated with the risk estimates.
xx
l 0 3:
l Interfacing Systems LOCA A major concern resulting from the BNL review of the interfacing systems LOCA analysis - in PLG-0465 and the RMEPS related to the determination of initiator frequencies. The effect of changing the initiator frequencies was determined by propagating the changes through the appropriate event trees in the RMEPS. The revised initiator frequencies resulted in the following changes to the frequencies of release categories S1W and S7W.
Mean Frequency Per Reactor Year Release Category PLG-0465 BNL Review S1W 4.0x10 8 1.4x10 7 S7W 6.3x10-8 1.1x10 6 The above changes in release category frequencies have no impact on indi-vidual risk of early fatalities if no evacuation or 1 mile evacuation is assumed. This is because release category S2W dominates this risk measure, and it has a frequency of 2x10 5 Only when a 2 mile evacuation is assumed (and the early fatality risk for category S2W becomes zero) do the above changes in release category frequencies change the original PLG-0465 estimates. However, with a 2 mile evacuation the early fatality risk is very low and well below the safety goal. The 200-rem dose versus distance curve in Figure S.3 is also not influenced by the above changes in release category frequency. This is because only release category S1W has a significant probability of exceeding a 200-rem dose, and the revised probability of this category is not sufficiently high for it to appear in Figure S.3.
There is of course uncertainty associated with predicting the frequency of interfacing systems LOCAs. However, the frequency of interfacing systems LOCAs resulting in release category S1W would have to increase by two orders of magnitude before the Seabrook dose versus distance curves would approach the curves given in NUREG-0396. One can therefore conclude that interfacing systems LOCA are unlikely to increase the risk profiles presented in PLG-0465 to the level presented in NUREG-0396. This is not too surprising because when no evacuation is assumed, the higher frequency events dominate risk and interfacing systems LOCAs did not contribute to the dose versus distance curves taken from NUREG-0396 (refer to Figure S.3.).
Accidents During Shutdown This topic was not originally addressed in PLG-0465 and a detailed assessment of such events is beyond the scope of the current BNL work on this project. However, the applicant was requested to provide information on the risk associated with accidents during shutdown. The results of the appli-cant's assessment of such accidents were presented in the form of sensitivity studies in a draft version of this report. The applicant provided additional frequencies to the existing release category frequencies given in Table S.1 to assess the impact on risk from accidents during shutdown. A base case and a bounding case were presented by the applicant. The additional frequencies associated with these accidents are given b'elow:
xxi
4 Mean Frequency'Per Reactor Year Release Category -
Power Base Case Bounding Operation Events Shutdown Events Shutdown Events S.5 1.1x10 " 1.7x10 5
.S.2 -2.1x10-5 , 4.9x10 7 S.6 6.5x10 7 ' ' 7.1x10 8 5x10 6 BNL was not in a position to assess the above frequencies for these events because there remained fundamental questions .regarding the modeling of these scenarios. However, in spite. of this the applicant's results were included in the draft report for comparison. with the BNL sensitivity study results on.other topics. It should be noted that the applicant considered the upper bound estimates to be very conservative. In ' particular, in order to assess the impact of these events, they were included in source term categories derived for accidents from full power, which could lead to predicts of shorter times and larger quantities of fission product release than would be expected from accidents during shutdown. -
In a subsequent submittal by the applicant, the consequences of accidents from -shutdown were revised. The applicant felt that 94 percent 1of accidents at shutdown would occur:at times later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after scram. Thus, the consequence estimates were reanalyzed assuming release times of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
The later release times resulted in dose versus distance curves which fall off at much shorter distances from the site boundary than the original dose versus distance curves. BNL has checked' this . result and confirmed that if the release does occur at times greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, then -the new dose versus distance curves are reasonable.
The results of the latest applicant's assessment of accidents at shutdown are reproduced in Figure S.S. As noted above, a detailed assessment of such events is beyond the scope of the current BNL review. However, based on our limited review of the applicant's assessment of these events, we still have reservations about the results. These reservations are discussed in greater detail in the body of this report but until they are resolved, we are unable to assess the validity of the risk estimates presented by the applicant in Figure S.5.
Induced Steam Generator Tube Rupture For accidents in which the primary system is at high pressure during core uncovery and melting, it is possible that large natural circulation flow patterns could develop within the primary system. These flow patterns could in turn heat-up and degrade regions of the primary system remote from the reactor core. Of particular concern is the possibility of degrading the steam generator tubes such that the primary system. could become open to the secon-dary system. If the secondary system were in turn open to atmosphere, then a direct path could exist between the primary system and the atmosphere, which bypasses containment. This topic was not included as part of the work scope for the current BNL review. However, the topic has been reviewed in detail by the NRC staff and is the subject of continuing NRC and industry research activities. Therefore, BNL performed sinple sensitivity studies to assess the xxii
potential impact of induced steam generator tube rupture on risk at Seabrook.
The results of the sensitivity study are given in Figure S.6.
Because of uncertainty in predicting events of this nature and the fact that BNL did not evaluate this issue in detLil, we were not able to develop a best-estimate frequency for induced SGTR. The sensitivity study therefore presents a range of possible frequencies for induced SGTR. The frequency of high pressure squences and the conditional probabilities of failure of the operators to decressurize and induced SGTR that were used in the sensitivity study are given below:
4.0x10 5 x 0.5 x 0.3 = 6.0x10 6 per reactor year 4.0x10 5 x 0.2 x 0.01 = 8.0x10 8 per reactor year.
In order to estimate the impact of the above probabilities on risk, an appropriate source term category had to be selected. It was decided to allo-cate SGTR events to release category S1W, which represents a large early bypass of the containment. It was felt that this was a conservative assump-tion because significant retention of the fission products in the secondary side could occur and this was not considered when calculating the SIW release fractions. The impact of adding the above frequencies to source term category S1W is illustrated in Figure S.6 The lower estimate of the frequency of induced SGTR has no impact on the risk estimates presented in Figures S.1-S.3. The higher estimate of the fre-quency of induced SGTR has no influence on the individual risk of early fatal-ities within 1 mile of the site boundary if no evacuation is assumed but does influence the 200-rem dose versus distance curves as shown in Figure 5.6.
Allocating the probabilities of SGTR events to release category S1W has the largest impact on the dose versus distance curves (refer to Figure S.4). How-ever,-the impact on the risk of early fatalities within 1 mile is negligible because S1W has very little risk of fatalities within this distance (refer to Table S.2). If the probabilities of SGTR events were added to release cate-gory S6W, the impact on the dose versus distance curves would be less but the risk of fatalities within 1 mile would increase slightly if no evacuation is assumed.
It should be noted that the range of frequencies used for the induced SGTR sensitivity study were developed to cover our lack of understanding in this area and that the NRC staff believes that the actual probability of a SGTR is closer to the lower estimate. However, one reviewer of the BNL draft report felt SGTR to be a potentially more "significant" issue than was implied in our evaluation. It was not BNL's intention in the draft report to minimize the potential importance of this issue, and the range we presented did not represent an upper bound. It was an attempt to reflect the best judgments of several experts on a very difficult subject. There is a great deal of uncer-tainty associated $.ith predicting such events and it is therefore prudent to indicate the impact on risk of a range of assumptions.
~
Containment Isolation Failure and Pre-existing Leakage The applicant's assessment of pre-existing leakage and containment isola-tion failure was reviewed by the NRC staff. Based on the NRC staff review of the information available, it was concluded that the purge and vent valves in xxiii
a fully closed configuration should provide reliable isolation of the Seabrook containment under severe accident conditions up to the pressure corresponding to 1 percent hoop strain in the containment.
The NRC staff also concluded that the applicant has presented a reasonable approach for the consideration of pre-existing leaks, both small and large. The approach ad)pted by the applicant was to use information on containment unavailability developed in a study by the Pacific Northwest Laboratory (PNL) to assess the impact on risk of pre-existing leakage. The applicant used this information to bound the effects of the data in the PNL study (NUREG/CR-4220) even though they considered that it did not apply to Seabrook. The results of the applicant's assessment are given in Figure S.7.
From an inspection of Figure S.7, it is apparent that the impact of the NUREG/CR-4220 data on the dose versus distance curves is not great. (
Containment Structural Capacity Based on its nonlinear finite element analysis of the Seabrook contain-ment, BNL concluded that a shear failure at the base of the cylindrical wall is a potential failure mode but would not occur before reaching a pressure of _
165'psig.
BNL agrees that the containment structure would reach a general yield (
state in the hoop reinforcing steel at a pressure of 157 psig and that it is appropriate to consider this pressure as a lower bound pressure for the hoop mode of failure. However, BNL believes that the median hoop failure pressure should correspond to the one percent strain level in the hoop reinforcing steel, which is a pressure of 175 psig. The above pressures are for the wet containment conditions. For the dry containment conditions the corresponding median failure pressure is 158 psig and the lower bound pressure (general yield) is estimated to be 145 psig. This latter value is based on the reduc-tion factor recommendation in Section 11.3.4.1 of PLG-0300.
With regard to containment penetrations, BNL believes that the failure pressures should be based on containment deformations assuming no bond strength between the reinforcing steel and concrete. Based on this assumption RNL estimates median failure pressures for the wet containment condition of 159 psig and 167 psig for two critical penetrations. For the penetration with the lower failure pressure, BNL agrees that a Type A (less than 6 square inches) leak path is appropriate for the median estimate; however a Type B (6 square inches to about 0.5 square foot) leak path should be considered as an upper bound estimate. For the penetration with the higher failure pressure, BNL agrees that a Type B leak path is appropriate for the medium estimate; however, a Type C (greater than 0.5 square foot) should be considered as an upper bound estimate.
For the dry containment conditions, BNL estimated the median failure pressures for the above two critical penetrations to be 147 psig and 152 psig, respectively. These values are also based on the reduction factor recommended in Section 11.3.4.1 of PLG-0300.
Although BNL has performed some independent calculations to support its conclusions regarding the containment stren'gth, it also relied on the results of calculations performed by PSNH and its contractors. Therefore, BNL xxiv
recommends that a complete and independent check of all relevant containnent strength calculations he performed by PSNH. PSNH committed to such a check in their letter to the NRC dated October 31, 1986 and has indicated that such a check has been completed.
Containment loads BNL's assessment of the capacity of the Seabrook containment (described above) has to be combined with severe accident loads (pressure / temperature histories) to determine the potential for early containment failure. BNL does not have Seabrook-specific containment loads and was not able to generate such loads given the limited scope of the current review. However, BNL has been involved in updating (NUREG/CR-4551, Volune 5) the risk profile for the Zion plant for input to the NRC's " Reactor Risk Reference Document," NUREG-1150.
The updating of risk for Zion was based on a methodology developed as part of the Severe Accident Risk Reduction Program (NUREG/CR-4551, Volumes 1-4) at Sandia National Laboratory (SNL). This methodology used expert judgment in an attempt to estimate the uncertainty associated with determining containment loads. The methodology was developed at SNL specifically for the Surry plant but was extrapolated to Zion at BNL. The Zion plant is very similar to Seabrook in terms of the containment volume to reactor power ratio. Thus, extrapolation of the Zion loads to Seabrook would give some indication of the impact of applying this new methodology to Seabrook. It must be emphasiZPd that this exercise should in no way be interpreted as a Seabrook-specific cal-culation. It simply gives some indication of the sensitivity of the Seabrook, results to the types of uncertainty in estimating containment loads discussed in NUREG-1150. It should also be noted that this work is preliminary and has not yet undergone full peer review outside of NRC and its contractors. It is, therefore, subject to revision.
The range of containment loads reported in Volume 5 of NUREG/CR-4551 for Zion is very wide and far exceeds the loads that would be considered credible by the applicant for Seabrook. Of particular interest is the loads at the time of reactor pressure vessel failure. These loads can range from about 60 psia to 200 psia depending on whether core melt is occurring with the primary system at high or low pressure and on whether or not containment heat removal systems, CHRS (sprays and fan coolers) are operating. The higher containment loads are postulated to occur for accidents in which the primary systen pres-sure remains high immediately before reactor pressure vessel failure. For these accidents, direct heating of the containment atmosphere by core debris or hydrogen combustion with a steam spike at the time of reactor vessel fail-ure are possible mechanisms for failing the containment. The applicant has presented information which indicates that these mechanisms are not credible ways of failing the Seabrook containment. However, as noted above, BNL does not have Seabrook-specific containment loads so we cannot, at this time, elim-inate these mechanisms as potential ways of failing the Seabrook containment.
For accidents with the primary system at high pressure and without the CHRS operating an approximate median load of 135 psia (120 psig) was predicted for Zion. If this median load is compared against the capacity of the Sea-brook containment given by the BNL review, one would conclude that the poten-tial for early containment failure at Seabrook is very low and would not influence the risk estimate in Figures S.1-S.3. However, the range of loads estimated for Zion implies censiderable uncertainty. The 95th percentile xxv
estimate of the probability of early containment failure at Zion is quoted as 0.17 in Volume 5 of NUREG/CR 4551. If this early containment failure proba-bility were also true for Seabrook, the risk estimates in Figures S.1-S.3 would increase significantly. However, the capacity of the Seabrook contain-ment is greater than Zion (the general yield for Seabrook is 157 psig compared with 134 psig for Zion) so the 95th percentile estimate of early containment f ailure should be lower at Seabrook than Zion. However, BNL cannot at this time quantify how much lower because we have not quantified Seabrook-specific containment event trees with Seabrook-specific containment loads combined with our estimate of the structural capability of the Seabrook containment.
Source Terms The fission product source terns.used in PLG-0465 were reviewed in terms of their consistency with the approaches used in WASH-1400 and found to be appropriate. A misprint in PLG-0465 related to the release of noble gases for release category S2W was discovered. However, correcting the noble gases release was found to have no impact on the risk profiles in PLG-046b. In addition, the argument presented by the applicant that water in the residual heat renoval (RHR) vault is sufficiently subcooled to warrant consideration of significant decontamination was found to be reasonable. This is an important consideration for the subset of interfacing systems LOCAs where the break location in the RHR line is low in the RHR vault. Under these circumstances, with the break location submerged considerable scrubbing of the aerosol fission products would occur. This would result in much lower aerosol fission product release than for accidents in which the break location was uncovered.
At the start of this section, we noted that the applicant considers the WASH-1400 source terms used in PLG-0465 to he very conservative and the appli-cant has high confidence that the source terms would not be exceeded in a real accident. BNL found the source terms used in PLG-0465 to be consistent with WASH-1400 methodology but we are not as confident as the applicant that they could not be exceeded. The new source term methods (refer to NUREG/CR-4551, Volumes 1-5) indicate that if the containment fails late or if there is gradual leakage from containment then the aerosol fission product release is likely to be lower than would be predicted by WASH-1400 methods. This is because WASH-1400 methods underpredicted aerosol agglomeration and settling.
Therefore, if the new methods were applied to release categories S2W and S6W, the predicted aerosol release would be lower than WASH-1400 values. However, the new methods also indicate that if containment fails early and the opening is large, then there is considerable uncertainty associated with predicting fission product release. The uncertainty ranges associated with fission product release in NUREG/CR-4551 can, for certain accident sequences and early containment failure modes, exceed the WASH-1400 predictions. This uncertainty would principally affect the S1W release category at Seabrook.
Consequence Model Tha applicant used the CRACIT code for their consequence assessments in PLG-046*i. BNL compared CRACIT predictions of dose versus distance with pre-dictions from the MACCS code, which was developed at Sandia National Labora-tory (SNL) under NRC sponsorship. The comparison of the dose versus distance curves for the CRACIT and MACCS codes was reasonably good. Therefore, BNL feels that the dose versus distance modeling in PLG-0465 is fairly presented xxvi
and that the relatively 'small differences between CRACIT predictions -and those computed by BNL using MACCS -are explained by differences in modeling tech ,
'niques used in the two codes.
BNL could not check the risk of early fatalities reported in PLG-0465 because we did not have the population distribution around the ~Seabrook site.
Therefore, as BNL. had only CRACIT results for early fatalities, it was decided to use CRACIT results for both early fatality risk and dose versus -distances in the BNL sensitivity study. This was done simply so that we had consistency between the two risk measures and not (as implied by the applicant's review of -
-the draft BNL report) to present more " conservative" CRACIT results. We found that CRACIT in general predicted dose versus. distance curves that extended further than the MACCS code and in this sense CRACIT is more " conservative" '
than MACCS. However, we note that MACCS predicts more early health risk than l 'CRACIT and therefore the use of the CRACIT results is probably not "conserva-tive" for this risk measure. In addition, we are concerned about the CRACIT predictions of early fatality risk close to the . site boundary for release category S1W (refer'to Table S.2). We consider these CRACIT predictions to be lower th:r. truld be calculated using MACCS. Therefore, use of the CRACIT pre-dictions for early -fatality risk in the population within one mile of the site boundary could give a more favorable comparison with the safety goal than would have been achieved.using MACCS predictions.
- 4. Conclusions
- A major conclusion of PLG-0465 is that there is no significant frequency of exceeding 200. rem beyond 1.5 miles at Seabrook and therefore a significant
,~
reduction of the current 10 mile EPZ is warranted. In order to draw this con-
! clusion, the ' applicant must have high confidence that. source terms, which result .in a 200 rem dose beyond 1.5 miles (release categories S1W and S6W in j Figure S.4), have a very low frequency (refer to Table S.1). This in turn i implies that the applicant must have confidence in their plant model and in their ability to predict low frequency events with high confidence.
The objective of this technical evaluation was to assess the results and
, conclusions in PLG-0465. The evaluation made no attempt to reassess or validate the total ' risk profile at Seabrook. The current review was an evalu-ation of selected issues related to the potential for a large early release of radioactivity at the Seabrook Station. The conclusions of the BNL review for
, each of the selected issues are briefly given below:
- 1) A major concern resulting from the BNL review of the analysis of In-terfacing Systems LOCA in PLG-0465 and RMEPS related to the determination of initiator frequencies. This concern resulted in significantly higher frequen-
- cies for interfacing systems LOCA than in PLG-0465. However, the higher fre-quencies suggested by the BNL review did not influence the risk estimates in
- j. PLG-0465.
i
- 2) Accidents during shutdown were not originally addressed in PLG-0465 or RMEPS but the applicant did provide an assessment of these accidents as l
part of an information request. A bounding analysis provided by the applicant indicated that accidents during shutdown have the potential to significantly impact the risk estimates in PLG-0465. However, the applicant's best judgment j was that if these accidents occur, they would most likely occur days after xxvii e
i
, -+.--,,,-,,.-..,r -.+-.4..,, - - . , . - . , .--w.--._,.-,-.- - , . - . , . - , . . - ------ .,.,r,- ~ -e.w y.--,,3-,ww.y ,,-e-,.ww., ,.-
shutdown and thdir impact on the current Seabrook risk estimates would be min-imal. However, BNL has reservations about the applicant's analysis of acci-dents during shutdown and until these reservations are resolved BNL cannot assess the validity of the applicant's risk estimates for this class of acci-dents.
- 3) Sensitivity studies performed by BNL indicate that induced steam generator tube rupture is a potentially risk important issue for accidents in which-the primary system is at high pressure. This issue was not reviewed in detail by BNL and questions remain on whether or not induced steam generator tube rupture would occur in the event of a severe accident. BNL considers that it has not yet been demonstrated that this issue is not risk significant for Seabrook.
- 4) The potential for Containment Isolation Failure and Pre-existing Leakage at Seabrook was not reviewed in detail by BNL or the NRC Staff. The NRC Staff concluded that the purge and vent valves in a fully closed position should provide reliable isolation under severe accident conditions. Estimates made by the applicant using generic data for containment isolation failure (NUREG/CR-4220) showed that this issue has a small impact on risk. BNL has not assessed the validity of the applicant's risk estimates ' for isolation failures.
- 5) Based on its nonlinear finite element analysis of the Seabrook con-tainment, BNL concluded that a shear failure at the base of the cylindrical wall is a potential failure mode but that it would not occur before reaching a pressure of 165 psig. BNL agrees that the containment structure would reach a general yield state in the hoop reinforcing steel at a pressure of 157 psig and that it is appropriate to consider this pressure as a lower bound pressure for the hoop mode of failure. However, BNL -believes that the median hoop failure pressure should correspond to the one percent strain level in the hoop reinforcing steel, which is a pressure of 175 psig. The above pressures are for the wet containment conditions. For the dry containment conditions the corresponding median failure pressure is 158 psig and the lower bound pressure (general yield) is estimated to be 145 psig.
- 6) With regard to containment penetrations, BNL believes that the fail-ure pressures should be based on containment deformations assuming no bond strength between the reinforcing steel and concrete. Based on this assumption BNL estimates median failure pressures for the wet containment condition of 159 psig and 167 psig for two critical penetrations. For the penetration with the lower failure pressure, a median leakage area of 6 in 2 is appropriate.
However, for the penet[ation with the higher failure pressure a larger median leakage area of 72 in is appropriate. For the dry containment conditions, BNL estimated the median failure pressures for the above two critical penetra-tions to be 147 psig and 152 psig, respectively.
- 7) BNL did not develop Seabrook-specific containment loads given the scope of the current review. However, BNL did develop Zion-specific contain-ment loads as part of updating (NUREG/CR-4551, Volume 5) the Zion risk profile for input to NUREG-1150. As the Zion plant is similar to Seabrook, it was
! decided to use the Zion-specific loads to give some indication of the sensi-tivity of the Seabrook containment to the types of uncertainty in estimating containment loads identified in NUREG-1150. The range of loads reported in xxviii 1 ..
- - - - . , - , - - - - . . . - - - , - , . . _ , - , - , , - ---r
L NUREG/CR-4551 is very wide (60-200 psia) and far exceeds the loads that the applicant considers credible for Seabrook. However, if the median Zion load is compared with - the capacity of the Seabrook containment given by the BNL review,'the potential for early containment failure at Seabrook is predicted
. to be very low. However, the range of Zion loads implies considerable uncer-tainty in estimating the probability of early containment failure. - Most of this uncertainty is given by accidents ,in which the primary system pressure remains high immediately before vessel breach. For these accidents direct i heating of the containment atmosphere by the core debris or hydrogen combus-tion with a steam spike at the time of reactor vessel failure have been postu-lated as mechanisms which could fail the containment. The applicant ~ has
- presented information which indicates that these mechanisms are not credible ways of failing the Seabrook containment. However, as BNL has not developed. -
'Seabrook-specific containment loads , we cannot confirm that the uncertainty associated with predicting the probability of early containment failure at !
Seabrook is as low as that claimed by the applicant.
- 8) The fission product source terms used in PLG-0465 were reviewed in
- terms of their consistency with the approaches used in WASH-1400 and found to j be appropriate.
- 9) The applicant used the CRACIT code for their consequence -assessments in PLG-0465. BNL compared CRACIT predictions of dose versus distance with predictions from the MACCS code, which was developed at Sandia National Laboratory (SNL) under NRC sponsorship. The comparison of the dose versus
. distance curves for the CRACIT and MACCS codes was reasonably good. There-i fore,' BNL feels that the dose versus distance modeling in PLG-0465 is fairly i presented and that the relatively small differences between CRACIT predictions
- and those computed by BNL using MACCS are explained by differences in modeling
- techniques used in the two codes. BNL could not check the risk of early fatalities reported in PLG-0465 because we did not have the population dis-tribution around the Seabrook site. However, BNL has questioned the validity of the very low CRACIT predictions of early fatalities close to the Seabrook site for the high energy S1W release category.
Based on the results of our focused review of PLG-0465 and RMEPS, BNL has
! low confidence that release categories S1W and S6W (or their equivalent) have frequencies as low as those given in Table S.I. Therefore, BNL is, at this time, rather less confident than the applicant that there is no significant frequency of exceeding 200 rem beyond 1.5 miles at Seabrook. Further work is needed in the following areas before BNL could alter its confidence level in the Seabrook risk' estimates:
- 1) A detailed evaluation of accidents during shutdown;
- 2) Further evaluation of induced steam generator tube rupture;
- 3) An independent quantification of Seabrook-specific containment event i
trees with Seabrook-specific containment loads including the current i
BNL assessment of the structural capability of the Seabrook contain-
, ment;
- 4) An independent systematic search f'or all accidents that might lead to early loss of containment integrity.
i !
XXIX i I
10-2 5 BACKGROUND ACCIDENTAL FATALITY RISK
- (5 FATALITIES PER 10,000 POPULATION PER YEAR)
{10-3 iE S
$104 -
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$30-5 3 SAFETY GOAL (.001 TIMES O BACKGRCUND RISK)
U y go-6 _ PLG-0465 Ui
?
= V"
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5 10-7 -
8 j WITH 1 MILE g . EVACUATION N - RMEPS RESULTS E
2 10-8 WITH NO IMMEDIATE PROTECTIVE ACTIONS 4 . . N 10'9 U# /' lll llll
....a... ..
.. figure S.1 Comparison of Seabrook Station sensitivity results using WASH-1400 source term methodology with background, safety
' goal individual and Rf1EPS risk levels.
(Reproduced from PLG-0465, April 1986) 1 XXX
10-3 LEGEND
--- PLG-0465 E
d: WASH-1400 0 10-5 -
(MEDIAN RESULTS)
Y a
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- 10*9 I I
\ \ l \
l l 100 10 1
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10 10 EARLY FATALITIES 4
Figure S.2 Comparison Of median risk of early fatalities at
- Seabrook station for different errergency planning l options. (Reproduced from PLG-0465, April 1986) 6 l
l XXXi
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- 1 10 100 1,000 OlSTANCE (MILES) l Figure S.3 Comparison of Seabrook Station results in this study and RMEPS with NUREG-0396 - 200-rem and 50-nim whole body dose plots for no immediate protective actions. (Reproduced from PLG-0465. April 1986)
XXXii
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- Ingmdients for these curves 3
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- 1. Potentially realizable 1 o. -
Containment .2 8 ** **
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3 No protective action 5 Release Category i E' ~ for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 52W 4. Release categories S3W
! x EE ~
\Containment and S5W did not exceed I $ {j a 200 rem dose j :- ,= . leakage 50 in2 j g 0.001 ,- Release {ategory S6W .
4 % : .
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0.1 1 10 100 .
Distance (miles) i l
)
! Figure 5.4 200 rem dose versus distance curves for various failure modes assuming no l immediate protective action. (Reproduced from PLG-0465.)
J i
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~ Shutdown events assuming 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> llG -
release, PSNH letter (NYN-87-002) ~
\ -
lgg g 200 REM o oo, I i\ i A i i il i i e ii...I , , , , , , , ,
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1 i
Figure S.5 Comparison of 200 rem-dose versus distance curves i for conservative assumption of no credit for l
operator recovery of open equipment hatch (calcu-lations performed by PSNH).
xxxiv
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I c.001 I ' ' ' 'I I ' ' ' ' ' ' I ' ' ' ' ' ' ' '
- 1 10 100 1,000 DISTANCE (MILES)
Figure S.6 Comparison of BNL sensitivity studies with PLG-0465 i
and NUREG-0396. (200-rem plots with no insnediate protective actions.)
i e
XXXV
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I c.coi I ' l ' ' is'il i i iienei i 1 lo too 3,o0a OlSTANCE (MILES) i Figure S./ Comparison of 200 rem-dose versus distance curves
(' for conservative interpretation by PSNH of NUREG/CR-4220 data. (Calculations performed by PSNH.)
i i
xxxvi I
~
4 Table S.1 Summary of Release Category Frequency Uncertainty Distributions -
i Annual Frequency '
3 Release Category Lower Bound Median Point' UPper Sound 5th Percentile 50th Percentile han Estimate. 95th Percentile
} S1 - Early Containment Failure 1.5(-9) 1.5(-9) 4.0(-9)* 5.2(-9)* 1.5(-9) 52 - Early Containment Leakage 3.5(-7) 7.5(-6) 2.l(-5) 2.0(-5) 1.0(-4 ) -
j S3 - Late Containment j g Overpressurization 5.l(-5) 8.3(-5) 1.4(-4) 1.4(-4) 2.3( 4) i E l SS - Containment Intact ,5.5(-5) 7.7(-5) 4 1.1(-4) 1.l(-4) 1.8(-4)
I SR - Containment Purge Isolation Failure <10(-10) 1.5(-8) 6.5(-7) 3.2(-7) 4.4(-6) -
57 - RHR Pump Seal Bypass 9.8(-10) 4.5(-9) 6.3(-8) 3.9(-8) 1.4(-7)
- Mean influenced by right tail of distribution beyond the 95th percentile. '
NOTE: Exponential notation is indicated in abbreviated form; i.e.,1.5(-9) = 1.5 x 10-9 ;
E i
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,, --~w ,- ~,--- -~-e w- -
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Table S.2 Early Fatalities Conditional on a Release Occurring in the Population Around the Seabrook Station Site Boundary Mean Early Fatality Risk No Evacuation 1 Mile Evacuation 2 Mile Evacuation Release Category Within Within Within 1 Mile Total 1 Mile Total 1 Mile Total S1W 9 746 9 746 9 746 S2W 121 122 8 9 0 0
$6W 385 -
734 193 542 2 63 4 g e 9
xxxvitt
1-1
- 1. INTRODUCTION
1.1 Background
The Seabrook Station Probabilistic Safety Assessment (SSPSA)1 was com-plated by. Pickard, Lowe, and Garrick (PLG) Inc., for the Public Service ,
Company of New Hampshire (PSNH) and Yankee Atomic Electric Company in December 1983 and submitted to the Nuclear Regulatory Commission (NRC). The NRC staff and its supporting contractor initiated an in-depth review of sections of the SSPSA related to determining those accident sequences that could lead to core damage. However, this effort was terminated prior to completion of the review. 2 A separate contract was placed with Brookhaven National Laboratory (BNL) to perform a very limited 8 review of those portions of the SSPSA related to core meltdown phenomenology, containment response, and radiological source terms. The BNL review 3 did not include an assessment of the physical strength 4
of the Seabrook Containment.
The key results of the SSPSA I are given below:
The mean and median values of the uncertainty distribution for core melt frequency were found to be 2.3x10 4 and 1.9x10 4 events per
, reactor-year, respectively.
Both the societal and individual risk provisions of the NRC safety goals were met by wide margins; hence, the risk to public health and -
safety was estimated to be extremely small.
Of fferent risk factors were found to have different key contributors.
Interfacing systems LOCA events and, to a lesser extent, seismic-induced transient events were the principal contributors to early ,
health risk. The contributors to core melt frequency and latent health risk were made up of a large group of initiators, including loss of offsite power, transient events, fires, and seismic events.
The dominant contributors to core . melt frequency were support system faults, external events, and internal hazards that affected both the
1-2 * '
core cooling and containment heat removal systems. As a result, a major fraction of the core melt frequency, 73%, was associated with sequences in which long-term containment overpressurization was indi-cated, while only 1% was associated with early containment failure.
. In contrast with previous containment analysis, the timing of contain-ment overpressurization in the above sequences was found to be measured in units of days rather than hours.
A major result of the SSPSA was that interfacing system LOCA events were the principal contributors to early health risk. The results of the SSPSA were updated in the Seabrook Station Risk Management and Emergency Planning Study (RMEPS), PLG-0432" to account for new insights regarding radioactive release source terms and the progression of sequences involving loss of coolant events that bypass the containment.
The purpose of the RMEPS was to present the results of a technical evalu-ation of emergency planning options and other risk rianagement actions that were under consideration for the Seabrook Station. The principal focus of the study was the evaluation of the impact of various protective actions such as evacuation and sheltering to various radial distances from the plant site.
RMEPS rebaselined the SSPSA analysis of these risk-important areas specifical-ly to establish an updated assessment of the risk at Seabrook so that alterna-tive emergency planning options could be developed. Therefore, RMEPS focused on new data and engineering insights about the initiation and progression of sequences involving interfacing systems loss-of-coolant accidents (LOCAs) and on the results of experimental and analytical research that provide an enhanced basis for assessment of radioactive material release (i.e., source
, terms) for a wider spectrum of accident sequences. Thus, RMEPS represents the appitcant's best estimate of risk at Seabrook.
(
l A second report related to emergency planning was pub 11shed 8(Seabrook i Station Emergency Planning Sensitivity Study, PLG 0465) which determined the radius of the Emergency Planning Zone (EPZ) that could be justified without consideration of any advances regarding the source term methodology since the
- completion of the Reactor Safety Study (WASH-1400)' in 1975. It is this i
^
1-3 second study that 'is the focuh of the current BNL review although in order to review PLG-0465 BNL had to evaluate the results of the RMEPS and the SSPSA.
The principal conclusion of PLG-0465 was that an EPZ at the Seabrook Station of 1 mile radius or less is more justified in terms of its risk management effectiveness than the current 10-mile EPZ was justified by the results of NUREG-0396.7 This conclusion was based on the results of the PLG-0465 Sensitivity Study, which are reproduced in Figures 5.1-5.3. These results were constructed ,without accounting for any new insights about source
! terms since WASH-1400. The conclusion was based on the following observa-
- tions
l The individual risk of early fatalities in the population within 1 mile of the site boundary with no immediate protective actions is less -
than the NRC safety goal (refer to Figure 1.1). This individual risk is substantially less when a 1-m11e evacuation is assumed.
The risk of early fatalities with a 1-m11e evacuation is comparable to
- the WASH-1400 results, which assumed a 25-mile evacuation (refer to i
Figure 2.2). The Seabrook Station results for a 2-mile evacuation are
- substantially less than those for WASH-1400.
?
i The risk of radiological exposures for 1, 5, 50, and 200-rem whole body doses with no immediate protective actions is less at 1 mile than the corresponding NUREG-0396 results at 10 miles (refer to Figure
- 3.3).
t The above observations led to the statement in PLG-0465 that "there is no significant frequency of exceeding 200 rem beyond 1.5 miles in the Seabrook sensitivity results." PLG-04G5 identiffed the following three areas as being I
the most influential in calculating the Seabrook risk estimates:
i The effectiveness of the Seabrook Station primary containment to
! either remain intact or to maintain its fission product retention l capability for periods much longerithan required for even delayed, ad j hoc protective actions.
1-4 - *
. A more realistic assessment of the strength and failure modes of the Seabrook containment than was possible within the state-of-the-art of PRA when the RSS was completed.
. A more realistic treatment of the initiation and progression of inter-facing systems LOCA sequences.
Note that of the three areas identified above as being the most influen- ,
tial to the risk estimat,es in PLG-0465 only the first is Seabrook specific, namely the effectiveness of the Seabrook containment. The other two areas refer to improvements in methodology and would therefore apply equally as well to other nuclear power plant risk assessments.
1.2 Scope and Focus of Review l At the request of the NRC, the BNL technical evaluation initially focused i on the following areas in PLG-0465 and RMEPS:
- Interfacing systems LOCAS l - Containment function:
- Isolation failure
- Pre-existing leakage
. - Structural capacity l - Containment loads
.- Seabrook-specific WASH-1400 source terms
- Site consequence model.
However, during the review process additional areas that were outside of the scope of the original BNL review were identified as potentially important to risk at Seabrook. Two of these areas were considered sufficiently impor-l tant to request the applicant to provide addition 6l information on the risk associated with such events. The two areas identified were:
- accidents during shutdown
- potential for induced steam generator tube rupture.
I e
. . 1-5 These areas were not initially included in the original BNL review because in the past they were not found to be dominant risk contributors.
However, as the risk estimates in PLG-0465 and the RMEPS are relatively low.
events that were previously considered to be unimportant now have the poten-tial to influence the Seabrook risk estimates. Thus, simple sensitivity studies were performed by the applicant and BNL to assess the potential influ-ence of these events on the risk estimates presented in PLG-0465.
By including consideration of these two additional areas (in addition to the other areas included in the original scope of the BNL review), it should not be assumed that BNL has performed a detailed and systematic search for all events that might be important to risk at Seabrook. Such a search is beyond the scope of the current BNL review.
The current review should therefore be i
regarded as an evaluation of selected issues considered important to assessing the validity of the results and conclusions in PLG-0465 and the RMEPS. The review therefore focused on assessing ways in which the Seabrook containment may fall or be bypassed early during a severe core melt accident.
1.3. Approach to Review l
The approach taken by BNL was to perform sensitivity studies in selected areas to assess the impact on the results in PLG-0465 of the BNL review. The BNL sensitivity studies used the conditional risk indices provided in PLG-0465 (and supporting documentation) to assess how changes in the probability of accident sequences and containment failure modes would change the Seabrook risk estimates. The sensitivity studies calculated revised 200 rem-dose versus distance curves for comparison with those given in Figure 1.3 and revised estimates of individual risk of early fatalities for comparison with the information given in Figure 1.1.
l The dose versus distance curves in Figure 1.3 were constructed from dose l
versus distance curves (given in Figure 1.4) for each of the source terms
! developed in PLG-0465.
These curves were then multiplied by their respective
! probabilities (given in Table 1.1) and summed. The combined dose versus distance curve was then normalized to thee total core melt frequency. To be l consistent with the NUREG-0396 approach, which used median probabilities taken L
1-6 , ,
i from WASH-1400, Figure 1.3 was based on the median probabilities given in Table 1.1.
l The information on individual risk of early fatalities within 1 mile of the Seabrook site boundary given in Figure 1.1 is based on the conditional risk indices given in Table 1.2 for the various PLG-0465 source terms. The earlier fatality risks were nultiplied by the mean frequencies in Table S.I. l
, summed, and then divided by the population at risk. Mean frequencies were used for this risk measure to be consistent with the NRC safety goal.
l l
The BNL review used the information in Table S.2 and Figure S.4 to assess how changes in the probdility of accident sequences and failure modes (and !
hence the probabilities of the source terms in Table S.1) would change the risk estimates given in Figures 1.1 and 1.3. Note that Table S.2 also gives the total early fatality risk for each release category and that these were the only risk measures available to BNL when the first draft of this report was issued for review. Thus, the preliminary sensitivity studies in the draft report used total early fatality risk and tried to infer how changes in total risk might reflect changes in the early fatality risk within 1 mile of the i Seabrook site boundary. However, by comparing the early fatality risk within 1 mile of the site boundary with the tots 1 risk it is clear that virtually all l of the early fataity risk for release category S2W and more than half of the risk for release category S6W occurs within 1 mile. It is also clear that most of the risk of early fatalities for release category $1W occurs beyond 1 mile of the site boundary. Therefore, a 2-mile evacuation eliminates all early ft.tality risk for release category S2W and virtually all for release l 1
category $6W. However, a 2-mile evacuation has no impact on the early fatality risk for release category S1W. Thus, it can be misleading to use the total risk of early fatalities as an indicator of the early fatality risk within 1 mile of the site boundary and this led to some confusion in the I ear 11er draf t, which has been corrected in this final version of the report.
When mean or median probabilities are used, a range of probabilities is obviously implied and the safety goal specifically states that an attempt has to be made to quantify the uncertainty associated with risk estimates. The
, applicant considers the WASH-1400 source terms used in PLG-0465 to be very t
. 1-7 conservative and has a high confidence that the source terms would not be exceeded in a real accident. Therefore, in the opinion of the applicant, only uncertainty in the probabilities of the accident sequences and containment failure modes would impact the risk estimates in Figures S.1-S.3. The appli-cant's upper bound or 95th percentile frequencies, which include consideration of the above uncertainties, are given for each of release category in Table S.I.
The impact of the 95th percentile frequencies in Table S.1 on the risk estimates in Figures S.1-S.3 is not great. The leading contributor to the risk of early fatalities without evacuation in Figure S.1 is release category S2W. The mean frequency of release category $2W increases by a factor of 5 f f the 95th percentile value is used. Therefore, the early fatality risk without evacuation would increase by about a factor of 5 if the 95th percentile fre-quencies were used. However, if 1 mile evacuation is assumed, use of the 95th percentile frequencies would result in an early fatality risk below the safety goal. Also, release category 52W is the only contributor to the 200-rem dose versus distance curves in Figure S.3 and, as this release category has no significant probability of exceeding 200 rem beyond 2 miles (refer to Figure S.4), changing its probability would not significantly change the results in Figure S.3.
1.4 Organization of the Report The previous section identified the focus of the BNL review of PLG-0465 and indicated the limitations of the effort. The report is organized to address each of the areas discussed in Section 1.2. Initially, in Section 2, those portions of PLG-0465 and the RMEPS (PLG-0432) related to system failure are reviewed to determine the appropriateness of the frequencies of accident sequences that could lead to early loss of containment Integrity.
Section 3 reviews the ability of the Seabrook Station primary containment to withstand the very severe pressure / temperature loads associated with core meltdown accidents. This is a very important review because the applicant considers that the Seabrook containment has, a significantly greater capability for containing core meltdown accidents than a number of other large dry
1-8 . -
containments that have been reviewed by the NRC staff over the last several years.
The sensitivity of the conclusions in PLG-0432 to uncertainties in con-tainment loads (pressure / temperatures histories) and containment performance (based on the review in Section 3) is explored in Section 4. The source terms used in PLG-0432, which were based on RSS methodology, are reviewed in Section
- 5. Finally, in Section 6, the site consequence model and the risk calcula-tions presented in PLG-0432 are reviewed.
1.5 References ,
- 1. "Seabrook Station Probabilistic Safety Assessment," Fickard, Lowe and Garrick, Inc., PLG-0300, December 1983.
- 2. Garcia, A. A. , "A Review of the Seabrook Station Probabilistic Safety Assessment," Draft Report, Lawrence Livermore National Laboratory, dated December 12, 1984.
- 3. Khat-Rahbar, M., et al., "A Review of the Seabrook Station Probabilistic Safety Assessment: Containment Failure Modes and Radiological Source Terms," NUREG/CR-4540, February 1986. ,
4 "Seabrook Station Risk Management and Emergency Planning Study " PLG-0432, !
Deceinbec 1985.
, 5. "Seabrook Station Emergency Planning Sensitivity," PLG-0465 April 1986
- 6. U.S. Nuclear Regulatory Comission, " Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," WASH-1400, NUREG-75/014, October 1975.
- 7. Collins, H. E., et al., " Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light l
Water Nuclear Power Plants," prepared for the U.S. Nuclear Regulatory Com-mission, NUREG-0396, December 1978 l :
i e
i I
l l
1-9 10*2 8ACKGROUND ACCIDENTAL FATALITY RISK 10'3 -
(5 FATALITIES PER 10,000 POPULATION PER YEAR)
? -
b10 4 -
Y Y
u, h10.s -
5 O .:.5..:
4'. SAFETY GOAL 1.001 TIMES BACKGROUNO RISK) h.go.4 _. -
PLG-0465
- 1o'1 -
V .
g .
EVACUATION N'
@ go.s . .
[ WITH 1 MILE
- RMEPS RESULTS g .' WITH NO IMME01 ATE PROTECTIVE ACTIONS
,o., .
I I
Figure 1.1 Comparisen of Seabrook Station sensitivity results using WASH-1400 source term methodolony with background, safety goal individual and RMEPS risk levels. (Reproduced from PLG-0165, April 1986.)
1-10 * '
10*3 4 LEGENO
--- PLG-0465 N
WASH.1400
$' to*5 -
(MEDIAN RESULTS) e 5
=
a: -
E '% % NO IMME0 LATE PROTECTIVE N ACTIONS to'0 I~ N , N o 1. MILE N
- d N IEvAcVATION \
W N \
\
B h 1o 7 \
E \ \
- \
.g I
\
io ' -
\ \
- -- - - , [ 2 . EVACUAflON uii.E \ {
HMEPS RESUI.YS % \
OFF SCALE N \
to 9 1 i \ I \ l -
loo lo' 10 2 go 3 10 4
105 EAMLY FATALITIES Figure 1.2 Comsarison of median risk of early fatalities at Sea) rook Station for different emergency planning options. (ReproducedfromPLG-0465. April 1986.)
9 b,___.___________________.___._________________..._____
1-11 I .
~
e ia i aiig i i i ii iig i i i i i i s i.
~
~
NUREG-0396 OE . -----
PLG-0465 ga -
,g -
. . . . . . . . . . . . .. . . R ,s , ,sV <1s ,O R o<
@b 1 .
J, w SEA 8 ROOK (200 REM CURVE STATIO,N 0F
.: 2 SCALE) o u,
[g 0.1 --
~~
cU ~
@3 .
h- 7~'% -
m
- o. "d
\. .
2
\
o$ .\ 200 So REM EU wo \ REM \ -
\ j - ,
\
g 0,01 --
g k "'"
0" : i \ :
g 8< _ l -
Ea -
1 I -
l
\
@5g c, t - -
_\ l
i 10 100 1,000 OlSTANCE (Mil.ES)
Figure 1.3 Comparison of Seabrook Station results in this study and RMEPS with NUREG-0396 - 200-rem and 50-rem whole body dose plots for no'imediate protective actions. (Repro-duced from PLG-0465 April 1986.)
t
l
, 1.0 g i o
" : T D -
A L S -
/ aroe early containment failure 2o -- 0.1 .
r bypass (high energy) -
i$
50 5
~.
Release Category Ingredients for these curves l
' SIW 1. Potentially - realizable gg - .
Containment source terms l "E _ leakage 1.5 .2in 2. Site characteristics g7 .Ol .
~
- 3. No protective action i Release Category for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l E' S2W 4. Release categories S3W EE -
\Containment and S5W did not exceed I
U ~
L a 200 rem dose
[E
~
leakage 50 in 2 i e
[ 0.001 -
Release Qtegory S6W -
I :
E
~
E -
E a .0001 , , , , ,,,,, , , , , ,,,,, , , , , , , , ,
0.1 1 10 100 ;
Distance (miles) l l
l Figure 1.4 200-rem dose versus distance curves for various failure modes. l assuming no immediate protective action. (Reproduced from ;
PLG-0465.)
m
__m..
Table 1.1 Sununary of Release Category Frequency Uncertainty Distributions Annual Frequency Release Category Lower Bound Median g88" Point %er M Sth Percentile 50th Percentile Estimate 95th Percentile S1 - Early Containment Failure 1.5(-9) 1.5(-9) 4.0(-9)* 5.2(-9)* 1.5(-9) ,-
S2 - Early Containment Leakage 3.5(-7) 7.5(-6) 2.1(-5) 2.0(-5) 1.0(-4) 53 - Late Containment Overpressurization 5.1(-5) 8.3(-5) 1.4(-4) 1.4(-4) 2.3(-4).
SS - Containment Intact 5.5(-5) 7.7(-5) 1.1(-4 ) 1.1(-4 ) 1.8(-4') y
~
S6.- Containment Purge Isolation :
Failure <10(-10) 1.5(-8) 6.5(-7) 3.2(-7) 4.4(-6)
S7 - RHR Pump Seal Bypass 9.8(-10) 4.5(-9) 6.3(-8) 3.9(-8) 1.4(-7)
- Mean influenced by right tail of distribution beyond the 95th percentile.
NOTE: Exponential notation is indicated in abbreviated form; i.e., 1.5(-9) = 1.5'x 10-9
1 . -
Table 1.2 Risk of Early Fatalities in the Population Around the Seabrook Station Site Boundary ,
Mean Early Fatality Risk Release
' Category No Evacuation 1 Mile Evacuation 2 Mile Evacuation Within Within Within !
1 Mile Total 1 Mile Total 1 Mile Total r
S1W 9 746 9 746 9 746 S2W 121 122 8 9 0 0 S6W 385 734 193 542 2 63 e
t a
0
2-1
- 2. SYSTEM EVALUATION In this section, those positions of PLG-0465 1 and the RMEPS2(PLG--0432) related .to determining the frequencies of accident sequences leading to core melt is reviewed. However, the review made no attempt to reassess or validate the total core melt frequency at Seabrook. Such reassessment or validation would require an extensive review and requantification of the results reported in the SSPSA3 and this was not within the scope of the current BNL evalua-
, tion. The current review was -therefore an evaluation of ' selected issues related to the potential for a large early release of radioactivity at the Seabrook Station. .The original focus of the RNL review was therefore directed to assessing:
Interfacing systems LOCAs Containment function:
Isolation failure Pre-existing leakage.
4 However, during the review process the following types of questions guided the focus of. the BNL review to other areas that were not in the origi-nal work scope:
- 1) whether the basic assumptions of the submittal are adequate, and whether the analysis confronts all the important issues
- 2) whether, given the basic assumptions, the plant-specific modeling is self-consistent and complete within the range of issues addressed by the modeling
- 3) whether the scenarios actually modelled are properly quantified (including common cause considerations).
I An example of a Type 1 completeness question is discussed in Section 2.3, namely, whether a high-pressure melt scenario can lead to steam generator tube degradation and concomitant containment bypass. This is a fairly generic question, and one that was not in the original scope of the BNL review.
j However, it was considered to be potentially important to risk and therefore we thought that it should be considered. This question is presently being l studied by the nuclear industry and the NRC staff and their contractors. BNL 1
_.,_.._e. . _ - . -- _..
. ,. . _ - - . __ __ _ m .
2-2 ' '
b did not have the time to review the issue in detail so we simply performed a
- ' sensitivity study to give an indication of its potential importance to risk.
i An example of a Type 2 completeness question is whether the check valve between residual heat removal (RHR) suction and the refueling water storage tank (RWST) is likely to fail in scenarios wherein the RHR system is overpres-surized. The Seabrook study raises and discusses this question, but with ut
, clearly establishing why the check valve failure probability is low enough to warrant not exploring such scenarios. This particular question is emblematic of a family of such questions which cannot be exhaustively tallied within the i
scope of this ifmited review, but which bears on the submittal's conclusions.
Finally, a particular instance of a type 3 question is discussed below in Section 2.1, wherein the frequency of multiple check valve failures as derived for initiating events is challenged.
Since the present concern is with the frequency of a substantial release of radioactivity, this section focuses on several areas which bear directly on this question. The strength of containment is addressed in Section3. In this
, section, selected modes of containment bypass loss of containment isolation are discussed. These include classical interfacing systems LOCA, accidents during shutdown, failure of containment isolation, and induced steam generator j tube rupture during a severe core melt accident.
4 2.1 Interfacing System LOCA i
l 2.1.1 General 1
According to the Seabrook RMEPS 2 one of the principal contributors domi-nating early health risk--and one which has been subjected to extensive re-i analysis since the SSPSA3 --is an Interfacing Systems LOCA that bypasses con-
- tainment. From all the potential pathways through which an Interfacing
}. Systems LOCA (ISL) may occur, the study identified six lines as possible initiators for ISLs:
i i f
2-3
. Four lines -in the cold leg safety injection (Low Pressure Injection
[LPI)/ Residual Heat Removal [RHR] Loop Return lines)
Two lines in the suction side of the RHR system.
The corresponding initiator frequencies as well as the core damage fre- <
quency due to these initiators were obtained by PLG as a result of an enhanced and innovative Interfacing Systems LOCA analysis, which involved new treat-ments of various aspects of the accident.
The new treatments are listed below:
More complete modeling of valve failure modes New data on check valve failures 'versus leak size i .
More realistic treatment of dynamic pressure pulse
. Explicit modeling of RHR relief valves Quantification of RHR piping fragilities to overpressure Modeling of RHR pump seal leakage Operator actions to prevent melt considered Thermal hydraulic and source term factors mode' led using MAAP c' ode" Uncertainties quantified.
e The initiator frequencies obtained by the new approach are:
Cold Leg Safety Injection-Path'(VI) = 4.Sx10 6 event / reactor year RHR Suction Side Path (VS) <= 3.3x10 6 event / reactor year which yield a core damage frequency of CDISL.= 4.4x10 8 event / reactor year.
RNL has performed a limited review of the new analysis, compiled a number of questions and observations, and performed a sensitivity study on the initi-ator frequencies. The objective of this subsection is to provide the results of this limited review.
l 2.1.2 Other.ISL Paths BNL also perfortred a cursory survey for other potential containment bypassing pathways for ISL.- The BNL survey identified some pathways (e.g.,
l RHR lines to the RCS hot legs, latdown line, excess letdown line, etc.) which i
l l
L-
2-4 ' '
were ignored in the Seabrook study. Although there may be justification for ignoring these pathways, BNL believes that all the potential pathways should have been identi fied and the basis for the rejection of each such path documented.
2.1.3 ISL Initiator Frequencies The determination of the ISL initiator frequencies is one of the most important parts of the Seabrook RMEPS.2 This determination depends on the correct estimate of the frequencies of relevant failure modes of the valves in the various interfacing lines. These valve failure modes are:
. Disc rupture or gross leakage of series valves (check valves) in the LPI line:,
. Disc rupture or gross leakage of MOVs, failure of stem mounted limit switches, and disc failing open when indicating closed, in the RHR suction lines.
The approach applied for modeling of initiator frequencies in the Seabrook study is based on two " innovative" steps:
a) Separation of the check valve gross (reverse) leakage failure mode into
" gross reverse leakage" and " failure to reseat on demand" failure modes, j which were treated together in earlier data bases.
b) An analysis of data on check valve leakage frequency versus leak rate for check valves of the RCS/ECCS system boundary. This step resulted in applying a reduced check valve leakage failure frequency in the quantifi-cation of the initiator models. (A result which is challenged below.)
In the process of surveying data of the Nuclear Power Experience (NPE) l data base5 , no disc rupture events were identified by PLG for check valves or MOVs. The maximum observed leak rate was 200 gpm. Leak rates were estimated based on other available evidence: the rate of boron concentration change in
! the accumulators, rate of pressure increase in the accumulators, and similari-ty to other occurrences for which the leak rates were known. To estimate the total number of check valve hours, the information provided in NUREG/CR-1363 6 on the number of valves in the ECCS in various PWRs was used. PLG's total number of check valve-hours was 1.08x108 . To estimate the frequency of check valve f ailure to reseat on demand, two types of data were used: estimates i
l
2-5 from generic sources of failure data, and experiential data from eight U.S.
nuclear plants for which PLG performed plant specific PRAs.
2.1.3.1 Check valve failure frequencies Since the check valve failure frequencies play a crucial role in the ISL analysis, BNL performed a somewhat more detailed review of that part of the Seabrook study. As a consequence of the review process the following observa-tions are made:
a) After a successive screening process of check valve failure events (start-ing from a total of 692 events at both PWRs and BWRs), PLG limited its data base "to those events involving check valve leakage in the ECCS and RCS/ECCS system boundary of PWRs. These were judged to be the closest category to the initially seated and tested check valves modeled in the analyses." The final number of failure events were: 17 accumulator check
-valve failures and 4 ECCS/RCS interface check valve failures, b) To estimate the total number of check valve hours, PLG used the total population of check valves in the ECCS instead of the corresponding subset of check valves at the ECCS interfaces. This resulted in substantial and inappropriate overestimation of check valve hours and thus a substantial underestimation of the check valve failure rate (i.e., check valve fail-ures divided by check valve hours).
c) The correct exposure time for check valve failures is not merely the time when the plant is operating. For example, check valves in the RHR are almost continuously exposed to potentially degrading conditions (during cold shutdowns, as well). A correction factor for pressure exposure of interfacing lines should be considered separately, in calculating the
. in'.tiator frequencies.
d) In many cases, PLG estimated single check valve leak rates from accumula-tor inleakages. It must be recognized that the deduced leak rates from accumulator inleakages relate to two check valves in series, rather than leakage through a single check valve. At leakages through two check valves in series, the lest,-leaking valve dominates (the other valve may be even wide open).
e) The leak failure frequencies versus leak rate curve presented in the PLG study (reproduced in Figure 2.1) is only a first approximation for a more
2-6 ' '
precise leak failure frequency versus relative leak rate curve. In par-ticular, this curve pooled data involving a variety of check valve sizes.
A more sophisticated treatment would require knowledge of the size popula-tion of check valves at the interfacing pathways.
f) The largest leak rate in Figure 2.1 is of the order of 200 gpm, whereas the arena of interest ranges to 65,000 gpm. The " linear" extrapolation to higher rates is not necessarily justified. If the shape of the distribu-tion is Pareto, the linear extrapolation is in order. However, if it follows a Rayleigh distribution, the extrapolation is not correct (but conservative). Seabrook-specific considerations (valve sizes, designs) are not addressed in the analysis.
g) The initiator models implicitly assume that the leak tests of the valves
" discover" all failures and valves behave as new after each test. The study does not describe the relevant test processes and the expected "real" efficiency of these tests.
h) The report does not consider common cause failures. Such failures can indeed happen due to boron deposition, improper maintenance such as installation of improper components (gaskets, seats, or valve disks) which may fail almost immediately or at a later time.
In order to quantitatively- estimate the consequences of some of the above mentioned deficiencies, BNL performed an independent detailed reevaluation of relevant chbck valve failure data. The process was facilitated by the availa- '
bility of relevant failure events selected for an independent study of ISL at PWRs, which is presently ongoing at BNL for the NRC. Tables 2.1 and 2.2 present failure events for High Pressure / Low Pressure isolation check valves selected by BNL. Table 2.1 contains events for the check valve ' leakage" "
failure mode. Table 2.2 presents the events for the valve " failure to reclose upon demand" failure mode. Table 2.1 also includes data on the estimated leak flow rates. These latter data are obtained essentially with the same method as those of Table 3.8 of PLG-0432.
Comparing the number of failure events of Table 2.1 with that of Table 3.8 of PLG-0432, one finds that Table 2.1 centains more events (41) than Table 3.8 of PLG-0432 (21), which may be the result of a somewhat more efficient selection procedure at BNL. .
2-7 In order to see what is the sensitivity of the initiator frequency for the check valve failure rate, BNL selected the subset of accumulator check valve failure events (the majority of check valve failures 35 events, summa-rized in Table 2.4) for which the total time of exposure can be correctly determined.
The total time of exposure of accumulator check valves for all the PWRs' in the U.S. is calculated in Table 2.3. The time from start of commercial operation of individual plants was used as " time of exposure" for these check valves, since water with boric acid constantly degrades these valves. The total number of check valve-hours obtained is 2.34x107.
Based on BNL-gathered data, the frequencies of accumulator check valve leakage events for various leak rate ranges are given in Table 2.4. The corresponding frequency exceedance/hr values are plotted against the check valve leak - rates in Figure 2.1. For comparison, Figure 2.1 shows also the PLG data. The shape of the curve is almost identical with that of PLG, but shifted higher, by alnost one order of magnitude, due to the higher number of failure events identified and the more precise value for check valve-hours.
It is appropriate to mention here several precautions concerning the leakage. failure characteristics derived from accumulator check valve failure events.
. In the NPE data base 5 the majority of interfacing check valve leakage events involve accumulator valves. Although this seeming bias could arise from the extra monitoring of the accumulator, it could also reflect a particularly severe environment acting on the valves. If the latter is true, then leakage exceedance frequency data (ordinates in Figure 2.1) may lead to overestimates of the frequency for other interfacing check valves.
l .
The leak flow rate data (" leak sizes"; abscissas in Figure 2.1) repre-sent lower limits for these quantities, because leakage flow rates l estimated from accumulator inleakages involve, in most of the cases,
! leakage through two check valves in series.
l As a result of these factors, a more realistic leakage failure l
i
. . = . -- - . . , _ _ _ - - -
2-8 exceedance frequency /hr versus leak rate curve for non accumu'l ator interfacing paths' may be sorrewhat . lower in frequency at lcw leak' rates, but might fall off more slowly with increasing leak rate than do the curves in Figure 2.1.
Creation of this more realistic curve ~ was beyond the scope of the BNL effort.
Since there are no more accurate data available, BNL recalculated the initia-tor frequencies by using the data obtained - for accumulator check valves.
Since the purpose of this calculation is to contrast the result with that of the PLG analysis (to see .the sensitivity of the initiator frequencies for-check valve failure rate), the same extrapolation and calculational techniques are used as those of PLG.
' 2.1.3.2 Cold leg safety injection path frequency 4
This section presents a revised estimate of the initiator frequency of interfacing LOCA through the injection lines. The calculation presented below is- intended to follow the PLG analysis step by step, except that the check valve failure statistics have been modified as indicated above. Subsequently, these modified initiator frequencies are propagated through the PLG model to illustrate new plant damage state frequencies. The following is, then, a recalculation of the PLG result using PLG methods but modifying the single
- check valve failure rate as previously discussed.
i From Figure 2.1, the median frequency of a single check valve failure resulting in leakage that exceeds the capacity of one charging pump (i.e., 150 gpm) is about 1.1x10 7 per hour. Assuming a lognormal distribution for this frequency and a range factor of 10 (which may be too conservative for this increased statistic) yields:
Frequency of Check Valve Failure Parameter (Leakage 150 gpm) 95th percentile 9.6x10 3/RY Mean 2.6x10-3/RY
- Median 9.6x10 4/RY l 5th percentile 6.6x10-5/RY I
, _ . . _ _.-_.__,-,-e.--._- .m, . , , ~ _ _ . _ . - . , _ _-.,.,r_...,_._,.-_..,_.m_,,_,,..s. . - - -
.m+
2-9 Similarly the median frequency of exceeding 1800 gpm is 1.4x10 s per hour.
Assuming a lognormal distribution with a range factor of 14 yields:
Frequency ofECheck Valve Failure Parameter (Leakage 1800 gpm) 95th percentile 1.7x10 3/RY Mean 4.4x10 "/RY Median 1.2x10 "/RY 5th percentile '8.8x10 6/RY The frequency of " failure to reclose on demand" for check valves, 4, is taken to be the same mean value as that used by PLG:
4 = 2.7x10 4/ demand.
By using Formula 3.14 of PLG-0432, the estimated mean frequency of failure of two series injection check valves, that produces leakage to the RHR system in excess of 150 gpm ist 4.90x10 5 events /RY.
Since there are four injection paths, the mean value for the Cold Leg Safety Injection Paths becomes VI = 1.96x10 " events /RY'.
Top event, LR in the injection path event tree (see Figure 3-4 in Reference 2) represents the fraction of the initiating event frequency, VI, in which the leakage not only exceeds 150 gpm, but also exceeds 1800 gpm. The product of LR and VI thus represents the dominant contributor to the frequency of overpressurization challenges to the RHR system due to failure of both check valves in the four injection paths. Based on the above values, LR has a mean value of .058 and the overpressurization frequency for the cold leg injection path becomes:
i i i 4) = 1.14x10 5/RY.
(0.058)(1.96t10 2.1.3.3 : RHR suction sideifrequench The same introductory remarks made at the beginning of Section 2.1.3.2 also apply here. That is, the following represents BNL's attempt to show the sensitivity of the PLG analysis to a modified check valve failure rate.
2-10 For an ISL to occur in the RHR hot leg suction path, failure of two series MOVs must occur. In the PLG-model for this path, the failure involves:
a) independent failures of both MOV valves, causing excessive leakage; or b) independent failure of one of the valves and a demand failure of the second valve, or c) " valve fails open while indicating closed", failure for the first valve and excessive leakage failure of the second valve.
In the PLG treatment, the frequency of M0V valve disc leakage and failure upon demand (due to a sudden pressure loading) were assumed to be identical to that for the check valves. For the frequency of failure of an MOV to close on demand but indicate closed, a mean value of Ad = 1.1x10 4 failure / demand was used in tne PLG treatment.
Applying the same approach as PLG (Formula 3.15 of PLG-0432) with the newly determined check valve leakage frequency, BNL recalculated the total (2 lines) suction side ISL frequency, VS. The new mean frequency for the RHR suction side path is:
VS = 1.44x10 4 events /RY. ,
The split fraction, LR, for the fraction of VS in which the leakage past the series M0Vs is greater than the capacity of the relief valves (see Figure 3.5 fr Reference 2), is practically the same as in the case of the cold leg injection lines (.058) and the overpressurization frequency for the RHR suction lines (again neglecting other insignificant contributions) becomes:
(0.058)(1.44x10 ") = 8.3Sx10 6/RY.
It is noted that for the BNL-calculated check valve leak rates, the PLG procedure of using the check valve leakage failure rates as " conservative" estimates for the leakage failure rates of the MOVs in the RHR suction lines is probably too conservative, and appropriate M0V leakage failure frequencies should be used if PLG redoes their analysis.
Furthermore, in the case of a PLG reanalysis of the RHR suction lines initiator frequency, VS, the following BNL observations should also be taken into account:
2-11 a) Inadvertent opening of the two MOVs due to common cause failures such as improper maintenance, malfunction of the interlock system, design error, improper tests..or testing operations.
b) Failure of the stem or other internal connections in valves equipped with limit switches or failure of a limit switch (including -improper mainte-nance such as reversing indication).
c) It is. difficult to see why only two MOVs have limit switches, instead of four.
d) It would be very useful to describe the valve inspections that are pro-mised each time the plant goes to cold shutdown or is refueled. For example, at a plant recently investigated by NRC, Region.1, everything was tested thoroughly, but the relays for the MOVs were not inspected.
e) Considerations should be given to operating procedures and the likelihood that the procedures will not be followed.
f) Interlock behavior.
2.1.4 Operator. Actions
! The ability of the Seabrook operators to diagnose, respond to, and miti- '
gate a Reactor Coolant System (RCS) to Residual Heat Removal (RHR) Interfacing Systems LOCA will be reviewed (without hardware reliability) in this subsection. Appropriate operator actions can mitigate the consequences of the ISL sequence that result in leakage outside containnent when the capacity of
- the RHR pump suction relief valves is exceeded and subsequent failure of the
!? RHR pump seals results. As discussed by PLG-0432,2 the success of these j, mitigative actions is dependent on the ability of the Seabrook operating staff
- based on their training and emergency procedures, to correctly diagnose a LOCA l outside containment. The correct diagnosis may be hampered by operator l confusion between symptoms associated with those LOCAs inside containment 3
which fill and pressurize the Pressurizer Relief Tank (PRT) by pressurizer relief or safety valve discharge flow and those associated with a RCS-RHR Interfacing Systems LOCA outside containment which also fills and pressurizes the PRT via the RHR suction relief valve discharge flow.
The following is the result of a brief. BNL evaluation of operator diagno-l sis and actions to mitigate the consequences of RHR pump seal failure
2-12 Interfacing Systems LOCA sequences at Seabrook and its assessment in PLG-0432,2 Section 3.1.4.3 entitled " Operator-Actions and Emergency Procedures."
This evaluation was preceded by an independent and fairly extensive familiarization preparation with the Seabrook procedures as they relate to the ISL to be studied. This preparation was followed by observation of a series of Seabrook Simulator demonstrated accident sequences which illustrated the distinguishing characteristics of the LOCA outside containment and the responses expected of the Seabrook operators. The BNL evaluation was per-formed by a former Senior Licensed Operator and Westinghouse Reactor Plant simulator Certified Engineer. A more complete evaluation of operator response would require a comprehensive Human Reliability Analysis (HRA) such as Team Enhanced Evaluation Method (TEEM)7 by a knowledgeable team of specialists providing expertise in PWR operations, PWR systems engineering and human reliability. This team would develop a detailed task sequence analysis of the Seabrook operating staff performing the detailed tasks required to mitigate these sequences and analyze the associated human reliability of the staff response using the analysis.
There are three sets of operator tasks identified by PLG-04322 which are to be important to the mitigation of the sequences by the Seabrook operating staff (each with a unique Operator-Action Sequence identification number in parenthesis),namely:
. Diagnose the RHR system LOCA (01)
. Isolate the RHR system LOCA (02)
. Provide makeup to the RWST (03).
To successfully accomplish these tasks, the operating staff must follow the appropriate parts of the following Seabrook procedures which are appli-cable to the RHR system LOCA event.
. Procedure E-0 (Reactor Trip or Safety Injection), Rev. 00, dated 05/16/86.
. Procedure E-1 (Loss of Reactor or Secondary Coolant), Rev. 00, dated 05/16/86. This procedure provides guidance for long-term cooling and stabilization. -
2-13
. Procedure ECA-1.1 (Loss of Emergency Coolant Reci rcul ation--ECR),
Rev. 00, dated 05/16/86. This procedure provides guidance for supply-ing adequate ECCS flow and plant stabilization.
. Procedure ECA-1.2 (LOCA Outside Containment), Rev. 00, dated 05/16/86. This procedure provides guidance on isolating the rupture.
Please note that ECA-1.2, Rev 00 needs to be revised to ensure that valves RH-V21 and -V22, t,he RHR pump discharge hot leg injection cross connec-tion valves, are closed prior to trying to identify and isolate a break in one of the low pressure systems. This need was identified by a detailed BNL review of the above four procedures.
The quantification of the three operator tasks identified by the Operator Action Sequence identification numbers 01, 02, and 03 above have been provided in PLG-0432,2 Table 3-10. According to the accompanying discussion in Section 3.1.4.3, "These operator actions include the hardware contribution, where applicable, and are based on enhanced procedures and instrumentation in order to aid the operators in their diagnosis of the event." For each of the three operator tasks, a " base" human error probability (HEP) with a "rrean" value of 0.005 has been identified as "0P".
This singular human reliability analysis HEP number is identified in PLG-2 0432 as ". . . recommended in Table 20-6 of NUREG/CR-12788 . . for follow-ing a procedure under abnormal conditions. This human error rate is inter-preted to have a mean value of 0.005 and to be represented by a lognormal distribution range factor of 10." Therefore, the only part of the three sets of operator tasks 01, 02, and 03 which changes their quantifications values is the hardware contribution since the human reliability quantification contribu-tion to each of these three operator tasks use the same HEP value of 0.005 with an error factor of 10. Each HEP is based on NUREG/CR-1278,a Table 20-6 (entitled " Estimated Human Error Probability (HEP) related to failure of administrative control"), Item (4) HEP (entitled "Use written operations procedures under abnormal operating conditions"). Therefore, no numerical differentiation is made to distinguish quantitatively among operator actions related to " diagnose," to " isolate," and to " provide." Even without using the
P-14 9
Human Reliability Analysis section (4.3) of NUREG/CR-2815 , more distinguish-able HEPs should have been selected.
The October 15, 1986 demonstration at the Seabrook Simulator with several relevant accident sequences and the abovementioned Seabrook abnormal / emergency related procedures provided (in the absence of a detailed TEEM 7 equivalent human reliability analysis performed on an actual Seabrook licensed operator shift) some reasonable assurance that a licensed Seabrook crew would adequate-ly perform the necessary, actions within the time required if the exact acci-dent sequence were programmed on the sinulator. This assurance is heightened especially since the Seabrook Training Center has recently instituted, in October 1986, a training module entitled "RHR Interface LOCA/ Student Handout,"
as part of its Requalification Training Program. The inclusion of such a module will reinforce the importance of the RCS-RHR Interfacing Systems LOCA.
Please note that this Figure 4 (entitled "RHR Interface LOCA Isolation Sequence") of this module confirms the need to revise Seabrook Procedure No.
ECA-1.2, Revision No. 00, dated 5/16/86 entitled "LOCA Outside Containment" to close (or verify closed) valves RH-V21 and -V22, the RHR pump discharge hot leg injection cross connection valves. This will allow the operators to identify and isolate a break in one of the low pressure systems.
Nevertheless, there were a number of concerns raised during a plant walk-through on the same date as the simulator demonstration which the Seabrook Sirsulator cannot adequately address. These concerns include the following:
a) Ability of RHR pump leakage to be detected in the control room - con-cern lies with vault compartmentation design, with the Equipment
! Vault sump not receiving leakage promptly thereby delaying level I detection input in the control room, l
b) Ability of RHR pump relief discharge into the PRT to be distinguish-able in the control room from the pressurizer relief and safety valve discharge - concern with the latter relief and safety valve dis-charge tailpipe temperatures.
In summary, the operator action analysis performed in PLG-0432,2 Section 3.1.4.3 appears to be superficial at besta The use of one single HEP value from one table of NUREG-12788 is an example of a lack of detailed and l
2-15 insufficient task analysis in assessing human performance appropriately. A 7
detailed TEEM equivalent human reliability analysis is a far more appropriate and rigorous approach to assessing Seabrook operator actions during a RCS-RHR Interfacing systems LOCA. Nevertheless, the simulator demonstration empha-sized a practical assessment of human reliability and task sequence timing.
In addition, the new training module and revised associated procedure ECA-1.2 reinforced the commitment to appropriately train the operator. Therefore, reasonable assurance that a licensed and trained Seabrook crew would adequate-ly perform the necessary, actions within the timeframe required will be pro-vided by the procedure change and training.
2.1.5 Break Location The " weakest link" of the RHR pressure boundary when subjected to acci-dental pressurization was identified by the applicant to be the RHR pump seals. A tabular listing of failure probabilities at 2250 psia showing pump seal failure probabilities ranging to 0.5 while metallic failure probabilities (piping, valves, and tubing) were 0.006 seems to support this observation.
The estimates of metallic component failure probabilities were based on:
a) accidental pressurization peak pressure limited to the initial RCS pres-sure of 2250 psia.
b) a probability of failure at the yield strength of the material to be 0.01 and the probability of failure at the ultimate strength of the material to be 0.99, c) the characterization of the overpressurization event as a quasi static process.
d) the statement that at 2250 psia, the stresses in the limiting RHR piping are only approaching yield stresses and the heat exchanger tube and other mechanical components are at a small fraction of their respective yield stresses.
The characterization of the overpressurization event as a quasi static process with a limiting peak pressure equal to the initial RCS pressure of 2250 is based on IDCOR evaluations which have not been reviewed by BNL. The assignment of 1% and 99% failure probabilities to the yield and ultimate
__p_._ _ - r
2-16 strengths of the material respectively is acceptable since a failure at yield is considered unlikely while a failure at ultimate is considered very likely.
The statement concerning the safety margins inherent in the RHR piping and metallic components and the basis for their calculation, including the influence of aging or time dependent effects on these safety margins was questioned during BNL's review. Of particular concern in this regard was the capacity of the potentially corrosion-degraded or embrittled heat exchanger tubes to withstand any dynamic loads associated with the overpressurization event.
The applicant responded to our question stating that the maximum pre-dicted stresses in the RHR heat exchanger tubes due to dynamic loads were approximately 50% of the yield stress value providing a margin of greater than 50% for tube thinning due to corrosion. They further pointed out that water chemistry is periodically sampled as part of the plant chemistry surveillance program thus minimizing the possibility of corrosive attack. Although we have not reviewed the applicant's calculations, it appears that documentation exists to support their failure probability estimates for the pressure boundary even when the effects of corrosion degradation are considered.
2.1.6 Event Tree Quantification This section summarizes the effect of observations made by BNL in previ-ous sections to the event tree quantification. One of main problems in the quantification of various ISL scenarios related to the determination of the initiator frequencies. The other observations and questions mainly expose the overall uncertainty of the frequencies of these accident scenarios.
The effect of the change in the initiator frequencies to the plant damage states can be demonstrated if the modified initiator frequencies, VI and VS, given in Sections 2.1.3.2 and 2.1.3.3 are propagated through the corresponding event trees. Table 2.5 presents the results of the BNL requantification. The table and its notation is essentially the same as Table 3-14 of the Seabrook EPZ Study. For convenience, in Table 2.5, the meaning of some plant dt. mage states has been repeated. From Table 2.6 the new value of the total core 2
~ _ . _ . . _ _
2-17 damage contribution due to ISL can be determined (the sum of PDS states 8C through IFV). This is:
. COISL = 1.37X10 6 event / reactor year.
The value obtained is much higher than the updated value (see Section 2.1.1) of the Seabrook EPZ Study. It is much closer to the result of an earlier assessment given in the SSPSA,3 which is, t
CDI st = 1.8x10 6 event / reactor year.
To summarize the limited BNL review of the Seabrook ISL analysis, it is our finding that the analysis, as reviewed, is not acceptable. BNL has shown that the check valve failure rates were underestimated resulting in over-optimistically low 'ISL initiator frequencies. The BNL sensitivity study was simply an attempt to demonstrate the effect of a more consistently calculated initiator frequency on the Seabrook analysis. This should not be taken to mean that BNL has accepted the remainder of the Seabrook analysis because only the initiator frequencies were changed in the BNL quantification. Having ,
found problems with the initiators, the remainder of the model was not reviewed to an equivalent depth; although, a number of comments concerning other parts of the analysis have also been offered. '
2.2 Accidents During Shutdown and Refueling Conditions The Seabrook RMEPS2 concentrated on accidents that would occur during power operation, and did not assess the risk during non-power operation. !
Table 2.6 lists the 6 modes of plant operations as defined in the Seabrook Technical Specifications. They are listed as follows: '
- Mode
- 1. Power Operation
- 2. Startup
- 3. Hot Standby
- 4. Hot Shutdown
- 5. Cold Shutdown -
- 6. Refueling.
v- - -- w , , ,--- n, . ,_ - -, -. -- ,-- ,,-n r_-,n _ - - - , , - - - - - - -~- --,-.e-
2-18 As far as early releases are concerned, there are some potentially significant contributors from operation in modes 4, 5, and 6. Seabrook Technical Specifi-cations do not address the status of containment isolation in mode 5, and re-quire isolation in mode 6 only during periods of fuel handling. Consequently.
it is possible to have a core melt accident with the containment wide open.
Based upon Seabrook's omission of risk during shutdown from their analy-sis, BNL turned to NSAC-84 to add perspective to the Seabrook review.
NSAC-84 10 is the only major study available that was performed specifically to assess the core damage frequency due to accidents during non-power operation at PWRs. It fa an innovative and detailed study for the Zion plant, using the plant-specific procedures and experience. Three types of initiating events were considered: loss of cooling, low temperature overpressurization, and loss of coolant. NSAC-84 results show that the dominant core damage sequences are due to loss of the RHR system and human errors. The contribution of LOCA to core damage frequency during shutdown and refueling is approximately ,
2x10 6/ calendar year. The contribution of low temperature overpressurization is < assessed to be less than 10 10/ calendar year. The total core damage frequency during shutdown or refueling was assessed to be 1.8x10 5/ calendar year which is comparable to the frequency of core damage at Zion during power operations, i.e., 5.7x10 5/ reactor year (internal events only). It was stated in the executive summary of NSAC-84 that "with the uncertainties involved, the risk of fuel damage during some period of a shutdown may be as great as the risk at power."
, RNL is involved in an on-going project to review NSAC-84 and investigate methods to improve the RHR capability of PWRs. BNL has found that extensive changes to NSAC-84 are required to correct its deficiencies, and the changes tend to increase the calculated core damage frequency. Four examples of the changes required are discussed here.
Example 1: NSAC-84 calculated the frequencies of station blackout during 3 types of outages: refueling, drained maintenance, and nondrained mainte-i nance. They are 8.23x10 5, 1.96x10 6, and 2.71x10 s per year respectively (Table C-27 of NSAC-84). Given a station blackout during an outage, core 1
- - . - , _ . - -- . - - , n--- _ , . , . - - - - - - - ,_
--- . - - - _ = _ - . _ - . - . . - - - - - _ . _ _ . - . . . - -
'2-19 damage is'almost surely to result if offsite power is not recovered. NSAC-84 did not include such sequences in the list of core damage sequences (Table 6-1 of NSAC-84). Inclusion of these sequences would have a significant impact on L the calculated CDF.
r
! Example 2:
MSAC-84 assumes that both RHR trains must fail in order to result t in the initiating event of " loss of shutdown cooling." Most of the time during shutdown or refueling, only one train is used and the standby train 4
will not start automatically. Therefore, loss of the operating train will lead to loss of cooling. NSAC-84 considers operator response to loss of i
cooling in the loss of cooling event tree, while loss of cooling is defined to i
be loss of both trains. This implies perfect operator response to the loss of '
- the operating train.
Example 3:
Based on NSAC-84, the dominant cause of loss of cooling is due to '
a spurious signal that isolates the suction line from the hot leg. It is also a dominant contributor to core damage. The evidence for such spurious signals .
4 is that three valve closures occurred due to unidentified causes in 27,888 1
hours. Using this evidence, strictly, the frequency can be estimated to be i 3/27,888 = 1.08x10 "/hr. Instead of using this frequency, NSAC-84 uses a Bayesian approach with an inappropriate prior ' distribution, i.e., a distribu-j i
tion that applies to mechanical failure of values and does not include the effects from unique sources 'of spurious control signals. The Bayesian f updating artificially reduced the frequency by approximately an order of magnitude.to 1.38x10-5/ hour.
- Example 4
NSAC-84 analysis of low temperature overpressurization may be too i optimistic. Events such as those of Turkey Point-4 11 indicate that the
] frequency _with which a rapid pressurization occurs with the RHR system i
isolated .and the PORVs unavailable is higher than 10 3 per year. The opera-j tors have only a few minutes to respond to the event before the pressure
{ reaches the setpoint for the safety valves. Therefore, the human error proba-l bility is not going to be very small. If the operators fail to terminate the overpressurization, the primary system pressure will reach the setpoint for I
the safety valves. At this pressure, the vessel rupture probability may be of i the order of 10 3 12-13 l
The following subsections discuss the possible initiating events from
! NSAC-84 which are also applicable for Seabrook.
L~
Operational experiences and
+-. . . _ , . _ _ _ . , _ _ _ . _ _ , _ _ _ _ _ . _ _ . , - _ _ _ _ _ . - _ _ , , _ _ _ _ __ , _ . . . _ - _ _ _ _ _ . _ . , _
2-20 ' '
causes of failures are provided for each type of initiating event. Some scenarios that may lead to core damage are provided based on reports in the related area. Related safety issues are also discussed.
2.2.1 Loss of Decay Heat Removal During Shutdown or Refueling As noted previously, the RHR system is designed to remove decay heat from the primary coolant system during modes 4, 5, and 6. Most of the time when the system is operating, only one train is actually running; the other train is either on standby or unavailable due to test or maintenance. If the operating train fails to continue running, the standby train will not start automatically. Therefore, loss of the operating train leads to loss of the system and operator actions will be required to restore it. The Office of Analysis and Evaluation of Operational Data (AE00) identified and analyzed 130 I
loss-of-DHR events at PWRs 14 during approximately 500 reactor years of opera-tions between 1976 and 1983. This analysis indicates that the situation involving loss-of-DHR systems is not improving. According to this experience base, the frequency of loss-of-DHR is estimated to be 0.25 per reactor year.
Table 2.7 lists the categories of the 130 events. It can be seen that automa-tic closure of suction valves and inadequate RCS inventory are the two domi-nant causes of loss of DHR. Automatic closures of suction valves were caused
- by spurious high pressure signals, loss of instrument bus, and human errors in-calibration of pressure transmitters. Inadequate RCS inventory was caused by human errors and inadequate vessel indications during drained-down opera-tions. It was estimated! " that approximately two-thirds of the events were human error related.
Upon loss of DHR, operators may be able to restore the failed train by reopening the spuriously closed suction valve, or starting the standby train.
, Alternative methods for decay heat removal include use of steam generators and use of charging purps. Typically several hours are available before core uncovery occurs. Therefore, it is very important that the operators must he able to recognize the loss of DHR. In the 130 loss-of-DHR events identified in tne AEOD study, the operators responded in a timely fashion, such that no serious damage resulted. However, the duration of, loss of DHR in some cases exceeded one hour. !
It was estimated " that if a loss of DHR occurs at ANO-2 1
r
2-21 (a- CE plant) 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after a reactor trip with the RCS in a partially drained condition, the onset of core uncovery may be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the loss-ofDHR. D.
C. Cook (a W plant) reported IS the results of a corresponding analysis that the onset of core uncovery would take place about one hour af ter the loss of the DHR system.
The dominant core damage sequences in NSAC-84 represent scenarios in which decay heat removal is lost and the operators fr.fi to determine that action to restore cooling is required. For example, the accident sequence with the highest frequency is a sequence in which the RHR suction valve is inadvertently closed, the operator fails to trip the RHR pumps, and also fails to determine that action to restore cooling is required. Its frequency is estimated to be 4.3x10 6 per reactor year. Such scenarios can be postulated for Seabrook. However, quantitative assessment of the accident scenarios must take into consideration the plant-specific information. For example, operator performance is strongly affected by the indications or alarms available in the control room. Zion does not have any alarm in the control room for inadver-tent closure of the RHR suction valves, while the Seabrook control room has an audible alarm on the video alarm system if the RHR pump is running with the suction valve closed. Another difference between Zion and Seabrook is that Zion has a single drop line and Seabrook has two drop lines. This is not expected to be a significant difference because the auto closure logic at Sea-brook will isolate both suction lines when a spurious signal is generated, i.e., a single pressure transmitter provides input to the interlock logic of the two inner isolation valves, and a separate pressure transmitter provides input to the interlock of both outer isolation valves. BNL performed a LER search for loss-ci-DHR events due to spurious closure of suction valves at plants with two drop lines. Seven events were found in approximately 40 reactor years. This indicates that the frequency of spurious closure of suction valves at plants with two drop lines is not any lower than the average frequency for all plants.
In response to the NRC request for additional information , the Public Service of New Hampshire (PSNH) provideo a shutdown risk analysis for Sea-brook 16 utilizing the results of NSAC-84. . Two differences between Seabrook and Zion were accounted for, viz., the number of hot leg suction lines, and
2-22 ' '
the support system interfaces with the RHR system. In the analysis, credit was taken for the additional suction line at Seabrook, based on the statement "For spurious valve closure to cause a loss of RHR cooling at Seabrook sta-tion, it is necessary to postulate either a conunon cause event involving one valve in each suction path, or a coincidence' of a single valve closure and maintenance being performed on the other RHR train." This reduced the frequency of loss of DHR by a multiplicative factor of 0.145 and the core damage frequency by approximately a factor of 2. From the information avail--
able to BNL, it is not clear that this reduction is justified. It is true that a single train of the RHR system is adequate for decay heat removal.
However, the standby train is not normally operating and will not start automatically when the operating train becomes unavailable. The analysis in reference 15 assumes perfect automatic start signals for the standby train or perfect operator response to the loss of the operating train; however, it has previously been shown that operator error is important in these sequences, and neglect of it here is inappropriate. In fact, operator actions to restore DHR is specifically modeled in the loss-of-cooling event tree. Also, taking credit for operator action in estimating tha initiating event frequency represents double counting of the operator.
The analysis of the support systems in reference 16 has not been reviewed by BNL. As was stated in the analysis, the differences in the support system interfaces with the RHR system are unfavorable for Seabrook. Therefore, it was judged to be unnecessary to review this in detail given the time constraints on the BNL review.
Two issues are related to the availability of RHR system, i.e., unre-solved safety issues A-45 and generic issue 99. A-45 addresses the adequacy of decay heat removal systems in existing light water reactor nuclear power plants. Generic issue 99 addresses the RHR suction line interlocks on PWRs.
BNL is currently involved in a project to investigate methods to improve the reliability of RHR systems during shutdown or refueling. The results of the project will be used towards resolution of generic issue 99. It is believed that these issues are applicable to Seabrook. In particular, the PSNH analy-sis 16 of shutdown risk for Seabrook is inadequate, if not incorrect, and requires many changes.
' o 2-23 2.2.2 Low Temperature Overpressurization Low temperature overpressurization may occur during shutdown as a result-of unanticipated addition of mass to the reactor coolant system, for example, inadvertent actuation of safety injection pumps, or imbalance of letdown and charging flows. Imbalance of letdown and charging flow may be caused by spurious isolation of the RHR system (thus negating letdown flow) or loss of instrument air that causes the letdown flow control valve to close and the charging line flow control valve to open. To protect the Seabrook plant against such scenarios, a low temperature overpressurization protection system is activated when the primary system is cooled down after a reactor trip. The system monitors the primary system pressure and temperature and actuates a main control board alarm when the pressure reaches a pre-determined fraction of the allowable pressure, and on a further increase in measured pressure, transmits an actuation signal to the PORVs and the PORV isolation valves.
Also, the safety injection pumps and one or more of the charging pumps are made inoperable during initial cooldown. In addition to the P0RVs, the relief valves in the RHR system may he available to relieve the pressure. Each RHR suction line has a relief valve with 900 gpm capacity at 450 psig, and each RHR discharge line has a relief valve with 20 gpm capacity at 600 psig. How-ever, these relief valves may be made ineffective if the RHR suction valves close automatically when the setpoint of 600 psig is reached, as was the case in the Turkey Point-4 events. Actually, the Seabrook Technical Specifications only require either both PORVs or both RHR suction relief valves to be avail-able. Seabrook Technical Specifications do not address the status of the pressure interlocks on the RHR suction valves when the PORVs are not avail-able. The PSNH comments 17 on the BNL draft version of this report claim that power to the RHR suction valves is removed when the RHR system is aligned for
! DHR.
BNL has not been given any documentation that would allow independent verification of this claim and in fact BNL does have documentation of the NRC disallowing a request by another utility to remove power to their RHR valves under the same circunstances. The BNL response to some of the PSNH comments is provided in Section 2.2.4 Two generic issues are related to the subject of low tenperature over-pressurization, generic issues 94 and 70 Generic issue 94 considers
o e 2-24 additional low-temperature-overpressurization protection for light water reac-tors. It has a "high" priority ranking.13 Enclosure 1 to reference 13 is the prioritization evaluation for the issue. It was stated in the evaluation that before 1979 30 events in PWRs were reported where the pressure / temperature of , ,
the reactor coolant system violated Technical Specifications. After 1979, following changes to operating procedures and the implementation of overpres-surization mitigation systems, there have been two reported events of over-pressure excursion events, i.e., the Turkey Point-4 events. Based on the operational experience and the use of the Vessel Integrity Simulation Analysis (VISA) code,la the prioritization evaluation estimated that the core damage frequency due to vessel rupture in a low-temperature-overpressurization event at Oconee 3 to be 4.5x10 8 per reactor year. Generic issue 70 considers the reliability of PORVs and their block valves. BNL is currently investigating the issue, and a draft report of the work is upcoming.
2.2.3 Loss of Coolant Accidents During Shutdown or Refueling NSAC-52 19 reviewed operating experience within 5 calendar years up to the end of 1981, and identified 10 loss of coolant events at PWRs. They were caused by the following causes:
- 1. Inadvertent manual initiation of RHRS supplied containment spray.
- 2. Inadvertent loss of inventory to the containment building sump and/or automatic initiation of recirculation mode of low pressure safety injection.
- 3. Inadvertent loss of inventory via the RHRS relief valves.
4 Inadvertent loss of inventory via mispositioned crossconnect or drain va'ves.
- 5. RHRS valve packing gland removal during plant pressurization, dislodging the valve packing and gland.
- 6. Gross valve packing leakage.
As for the loss-of-cooling initiating event, LOCA during shutdown or refueling requires operator response to terminate the inventory loss and to provide inventory make up. The NSAC-84 analysis for Zion assessed the core damage frequency due to a LOCA at shutdown or fueling to be approximately 2x10 6 per reactor year. The dominant scenario is that the operator fails to
2-25 close the RHR return valve to the RWST after draining the cavity, on reestab-lishing RHR flow, a LOCA via the RWST vent outside the plant occurs, and the operator fails to respond to it. BNL does not have the Seabrook procedures used during shutdown, and therefore cannot judge if the same sequence is applicable to Seabrook. However, the LOCA experience identified in NSAC-52 is applicable to Seabrook.
2.2.4 PSNH Comments on BNL Draft Report PSNH provided an analysis 16 that makes use of NSAC-M in response to BNL's finding that risk at shutdown had not been addressed. PSNH also provided some comments l7 on sections 2.2.1 to 2.2.3 of the draft version of this report. This section simply provides a response to some of the PSNH input with the intent to clarify some misquoting or misinterpretation of the BNL draft report by PSNH.
1.
The PSNH analysis borrowed sequences from NSAC-84 and inherited the prob-lems and mistakes of NSAC-84 such as those outlined in this report (begin-ning of Section 2.2).
2.
PSNH states "As noted by BNL, operator failure to restore the standby train was not explicitly included in the Seabrook analysis." BNL did not make such a statement. On the contrary, BNL believes that the Seabrook 1
analysis ' erroneously assumes perfect operator response to loss of the operating train of the RHR system.
3.
The PSNH analysis of spurious valve closure indicates a misunderstanding of NSAC-84. Operator response to loss of cooling due to spurious valve closure is explicitly modelled in the loss of cooling event tree of NSAC-84.
For example, top event RT represents operator tripping the operating pump, top event RH represents reopening RHR suction valves, and restarting at least one RHR pump.
The PSNH analysis 17 first considers operator response to spurious valve closure to reduce the frequency of loss of DHR. It then uses the loss of cooling event tree in NSAC-84 that considers operator restoration of RHR again to calculate the core damage frequency. This way the core damage frequency is artificially reduced by taking credit for operator action twice. .
~
s r 2-26 4.- The PSNH' response to the draft version of 'this report provided a statement.
concerning procedural removal of power to the RHR suction valves. 'Such procedures have not been made available to BNL. It is known to BNL that Diablo Canyon requested NRC permission to similarly remove power, and the NRC staff reviewed the RHR isolation valve operating procedures and found
~
that the licensee should retain power avai1able to the MOVs when the RHR
~
system is in operation.20 The concern is that if power is removed from the RHR MOVs to remove the possibility ~of an inadvertent closure, then no ready means would be available to isolate the RHR system should it rupture or develop a leak outside containment. Assuming such procedures are used at Seabrook, the whole shutdown risk analysis will have to be redone'.
because spurious isolation of the suction valves is the dominant contribu-tor to core damage, and the PSNH analysis does not reflect the fact that power is removed.-
- 5. PSNH provided some discussion on the consequences l7 of accidents during shutdown. BNL believes that it. is not very meaningful to consider the consequences before the analysis on core damage scenarios is in a good shape. The calculation 17 on the mean time of scenario ' initiation is cor-rect, assuming that the distribution of the time of. failure is uniform.
However, the distributfori of the time of failure is not uniform. A more accurate way to calculate the mean time of scenario intiation is to esti-mate the times of scenario initiation for all core damage scenarios and use the frequencies of the core damage scenarios as the weights to calcu-late the weighted average of the times of scenario initiation.
2.2.5 Summary of the Shutdown Risk Review The Seabrook analysis did not. originally address shutdown risk. In response to BNL questions forwarded by the NRC, Seabrook provided a shutdown risk analysis based upon selected modifications to NSAC-84 BNL was well along on its independent review of NSAC-84 as part of a separate project when the Seabrook response was received. Based upon the scope and direction of the BNL review as outlined at the beginning of Section 2.0, BNL concludes that the shutdown risk assessment by Seabrook is unacceptable as presently documented.
The bases for this conclusion are detailed in the preceding subsections and in
s- a 2-27 sunmary are that NSAC-84 has a number of deficiencies and that the modifica-tion of_NSAC-84 to represent Seabrook also has a number of , deficiencies.
2.3 Induced Steam Generator Tube Rupture (SGTR)
For accidents in which the primary system is at high pressure during core uncovery and melting, it is possible that large natural circulation flow pat-terns could develop within the primary system. These flow patterns could in turn heat up regions of the primary system remote from the reactor core. As the primary system heats-up, it is possible that parts of the pressure boun-dary could degrade. Of particular concern is the possibility of degrading the steam generator tubes such that the primary system could become open to the secondary system. If the secondary system were in turn open to atmosphere, then a direct path could exist between the primary system and the atmosphere, whicn bypasses containment.
This is a very important topic for review because it could potentially lead to a relatively large early release of radioactivity. The topic was not included as part of the work scope for the current BNL review. However, the topic was reviewed 21 by the NRC staff and is the subject of continuing NRC and industry research activities.
Scoping studies were performed to assess the impact of induced steam generator tube rupture on risk at Seabrook. First, the frequency of accidents in which the primary system would be at high pressure had to be determined.
The applicant estimated 22 the frequency of high pressure sequences in which a SGTR might have an effect to be 4x10 5 per reactor year. The NRC review 21 considered this estimate to be significant and therefore concluded that it needed further consideration. It was used as the basis for the BNL scoping study.
Given that core meltdown occurs with the primary system at high pressure, the probability that the steam generator tubes will fail had then to be deter-mined.
In addition, it is also possible (provided methods are available) for the operators to depressurize the primary systen prior to induced failure of
2-28 the SGT. The probability of successful depressurization had also to he deter-mined. .
Estimating the probabilities of the above events is subject to signifi-cant uncertainty. However, the Severe Accident Risk Reduction Program at SNL attempted to quantify these probabilities by use of expert judgment. The probabilities were developed specifically for the Surry plant and reported in Appendix B of NUREG/CR-4551, Volume 2.23 The experts concluded that there was a conditional probability of 0.8 for the operators to successfully depres-surize the primary system. The BNL review team considers the 0.8 probability rather optimistic given that procedures do not exist for this operation and therefore consider the conditional probability to be indeterminant (CP = 0.5) without further analysis. In addition, the experts felt that the probability of an induced steam generator tube rupture might be between 0.01 and 0.1 (for both snall and large tube ruptures) conditional on no depressurization. These estimates are reasonably consistent with an earlier NRC memorandum 24 on this subject, which suggested a conditional probability of about 0.01 to 0.3 for SGTR given a high pressure core meltdown. It was therefore decided by BNL to use a range of 0.2 to 0.5 for _the conditional probability of failure by the operators to depressurize and a range of 0.01 to 0.3 for induced SGTR to assess the impact of this phenomenon on risk 'at Seabrook. The results are sumnarized in Section 2.5.
, 2.4 Containment Isolation Failure The applicant's assessment of pre-existing leakage and containment isola-l tion failure was reviewed2 s by the NRC staff. Based on its review of the information available, the staff concluded that the purge and vent valves in a fully closed configuration should provide reliable isolation of the Seabrook l containment under severe accident conditions up to the pressure corresponding to 1 percent hoop strain in the containment.
I l
The staff also concluded that the applicant has presented a reasonable approach for the consideration of pre-existing leaks, both small and large.
The approach 16 adopted by the applicant was to use information on containment unavailability developed in a study 26 by the Pacific Northwest Laboratory
i . 2-29 (PNL) to assess the impact on risk of pre-existing leakage. The applicant used this information to bound the effects of the data . in the PNL study 26 (NilREG/CR-4220) even though they considered that it did not apply to Seabrook.
2.5 Summary
.In this section, the BNL findings related to each review area are briefly sumarized. Sensitivity studies have been performed using the applicant's conditional risk indices to show how the dose vs distance and risk profiles might change as a result of the concerns raised in this section. The summa-ries included a discussion on the uncertainties associated with the risk estimates.
Interfacing System LOCA A major concern resulting from the BNL review of the interfacing systems LOCA analysis in PLG-0465 and the RMEPS related to the determination of initiator frequencies. The effect of changing the initiator frequencies was determined by propagating the changes through the appropriate event trees in the RMEPS. The revised initiator frequencies resulted in the following changes to the frequencies of release categories S1W and S7W.
Mean Frequency Per Reactor Year Release Category PLG-0465 BNL Review S1W 4.0x10-8 1.4x10 7 S7W 6.3x10 8 1.1x10 6 s The above changes in release category frequencies have no impact.on indk-vidual risk of early fatalities within 1 mile of the site boundary if no
~
evacuation or 1 mile evacuation is assumed. This is because release 'categohy S2W dominates this risk measure, and it has a frequency of 2x10 5 Only when a 2 mile evacuation is assumed (and the early fatality risk for cat,egory S2W' becomes zero) do the above changes in release catqgory frequencies change the -
original PLG-0465 estimates. However, wi,th a 2 mile evacuation the early fatality risk is very low and well below the safety goal. The 200-rem dose I
l
, e 2-30 versus distance curve given in PLG-0465 is also not influenced by the above changes in release category frequency. This is because only release category S1W has a significant probability of exceeding a 200-rem dose, and the revised probability of this category is not sufficiently high for it to influence the 200-rem dose versus distance curve in PLG-0465.
There is of course uncertainty associated with predicting the frequency of interfacing systems LOCAs. However, the frequency of interfacing systems LOCAs resulting in release category SIW would have to' increase by two orders of magnitude before the Seabrook dose versus distance curves would approach the curves given in NUREG-0396. One can therefore conclude that interfacing systems LOCA are unlikely to increase the risk profiles presented in PLG-0465 to the level presented in NUREG-0396. This is not too surprising because when no evacuation is assumed, the higher frequency events dominate risk ard interfacing systems LOCAs did not contribute to the dose versus distance curves constructed in NUREG-0396.27 Accidents During Shutdown This topic was not originally addressed in PLG-0465 and a detailed assessment of such events is beyond the scope of the current BNL work on this project. However, the applicant was requested to provide information on the risk associated with accidents during shutdown. The results of the appli-
- cant's assessment of such accidents were presented in the form of sensitivity studies in a draft version of this report. The applicant provided additional frequencies to the existing release category frequencies to assess the impact on risk from accidents during shutdown. A base case and a bounding case were presented by the applicant. The additional frequencies associated with these accidents are given below
Mean Frequency Per Reactor Year
- Release Category l
Power Base Case Bounding Operation Events Shutdown Events Shutdown Events S.5 1.1x10 1.7x10-5 S.2 2.1x10-5 . 4.9x10-7 S.6 6.5x10 7 7.1x10 8 5x10 6
v ;
t i .
2-31
.+
t j BNLL. was
+ !
not in a position to asse'ss the above frequencies for these t
events. because there remained fundamental questions' regarding the modeling of these scenarios-. However, in spite of this, the applicant's 'results were included in -the BNL draft report "for comparison with the sensitivity study results on other topics. It should be noted that"the applicant considered the upper bound estimates to be very conservative. In particular, in order to 3 assess the impact of these events, they were included in source term cate-gories derived for accidents from full power, which could lead to predicts of shorter times and larger quantities of fission product release than would be s
expected from accidents during shutdown.
In a subsequent submittal l7 6 the applicant, the consequences of acci-dents from shutdown were revised. The applidant felt that 94 percent of acci-dents at shutdown would occur at times later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after scram. Thus, the consequence estimates were-reanalyzed assuming release times of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
The later releaseitines 'resulted in dose versus distance curves which fall off at much shorter distances froin the site boundary than the original dose versus s ,
distance curves. BNL has checked this result and confirmed that if the release does occur at times greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, then the new dose versus distance curves are reasonable.
. The results of the latest applicant's assessment of accidents at shutdown are reproduced in Figure 2.2. As noted- above, a detailed assessment of such
~
events. is, beyond the scope of the current BNL review. However, based on our limited review of the applicant's assessment of these events, we still have restryations .about the results. These reservations are discussed in greater detail in Section 2.2', but until they are resolved, we are unable to assess
.the. validity of the rist estimates presented by the applicant in Figure 2.2.
Induced Steam Generator Tube Rupture In Section 2.3 a sensitivity study was suggested to assess the impact of induced SGTR on risk , at Seabrook. The frequency of high pressure sequences
'taken together with the conditional probabilities of failure to depressurize
. . - - . ~ . - .- _. . - ,. . _. .- - - - .
'2-32 and induced SGTR given in Section 2.3 give.the following range of probabili-ties of induced SGTR:
4.0x10 8 x . 0.5 x 0.3 = 6.0x10 6 per reactor year 4
4.0x10 5 x 0.2 x 0.01 = 8.0x10 s per reactor' year.
In order to estimate the impact of the'above probabilities on risk, an appropriate source term category had to be selected. It was decided to allo-
~
i cate SGTR events to release category S1W, which represents a large early bypass of the containment. It was felt that this was a conservative assump-tion because significant retention of the fission products in the secondary
- side could occur and. this was not considered when calculating the S1W release fractions. The impact of adding the above frequencies to source term category
, S1W is illustrated in Figure 2.3.
The lower estimate of the frequency of induced SGTR has no impact on the i risk estimates presented in PLG-0465. The higher estimate of the frequency of
- induced SGTR has no. influence on the individual risk of early fatalities with-t in 1 mile of the site boundary if no evacuation is assumed but does influence the 200-rem dose versus distance curves as sh_own in Figure 2.3. Allocating the probabilities of SGTR events to release category SIW has the largest
! impact on the dose versus distance curves (refer to Figure 1.4). However, the impact on the risk of early fatalities within 1 mile is negligible because S1W has very little risk of fatalities within this distance (refer to Table 1.2).
If the probabilities of SGTR events were added to release category S6W, the impact on the dose versus distance curves would be less but the risk of fatal-ities within 1 mile would increase slightly if no evacuation is assumed.
It should be noted that the range of frequencies used for the induced l
SGTR sensitivity study were developed to cover our lack of understanding in this area and that the NRC staff believes that the actual probability of a SGTR is closer to the lower estimate. However, one reviewer 28 of the BNL
. draft report felt SGTR to be a potentially more "significant" issue that was implied in our evaluation. It was not BNL's intention in the draft report to minimize the potential importance of this . issue, and the range we presented did not represent an upper bound. It was an attempt to reflect the best
2-33 judgments of several experts on a very difficult subject. There is a great deal of uncertainty associated with predicting such events, it is prudent to indicate the impact on risk of a range of assumptions.
Containment Isolation Failure and Pre-existing Leakage This issue addresses the possibility that containment may not be isolated during or immediately following an accident. This area has not been reviewed in detail by BNL or the NRC Staff. The NRC staff concluded that the purge and i
vent valves in a fully closed position should provide reliable isolation under severe accident conditions. Estimates made by the Applicant using generic data for containment isolation failure (NUREG/CR-4220) are shown in Figure 2.4 and indicate that this issue has a small impact on risk. BNL has not assessed the validity of the applicant's risk estimates for isolation failures.
2.6 References
- 1. "Seabrook Station Emergency Planning Sensitivity Study," PLG-0465, April 1986.
- 2. "Seabrook Station Risk Management and Emergency Planning Study," PLG-0432, December 1985.
- 3. "Seabrook Station Probabilistic Safety Assessment," PLG-0300, December 1983.
- 4. "MAAP-Modular Accident Analysis Program Users Manual," Technical Report on IDCOR Tasks 16.2 and 16.3, May 1983.
- 5. " Nuclear Power Experience (NPE)," S. M. Stoller Corporation, updated monthly.
- 6. Hubble, W. H. and Miller, C., "Date Summaries of Licensee Event Reports I
of Valves at U.S. Commercial Nuclear Power Plants," NUREG/CR-1363, June 1980 7.
" Team-Enhanced Evaluation Method (TEEM) Procedures--An Enhanced Human Reliability Analysis Process," Informal Report BNL-38585, Rev.1. Decem-i ber 1986.
- 8. " Handbook of Human Reliability Analysis With Emphasis on Nuclear Power Plant Applications," Final Report NUREG/CR-1278, August 1983.
2-34 . *
- 9. "Probabilistic Safety Analysis Procedures Guide," NUREG/CR-2815, Rev.1 Vol. 1,_ August 1985.
- 10. " Zion Nuclear Plant Residual Heat Removal PRA," NSAC-84, July 1985.
- 11. "0verpressurization of Reactor Coolant System," IE Information Notice No. 82-17, U.S. NRC, June 11, 1982.
- 12. Burus, T. J., et al., " Pressurized Thermal Shock Evaluation of the Oconee-1 Nuclear Power Plant," NUREG/CR-3770, Draft, April 1984.
- 13. NRC Memorandum from H. R. Denton, Director, _ Office of NRR, to R. M.
Bernero, Di rector, Division of System Integration, on the Subject of Schedule for Resolving and Completing Generic Issue No. 94, July 23, 1985. ,
- 14. Ornstein, H., " Decay Heat Removal Problems at U.S. ,
Pressurized Water Reactors," Office for Analysis and Evaluation of Operational Data. U.S.
NRC, December 1985.
j 15. Indiana and Michigan Electric Company,. Licensee Event Report (LER)'
50-316/84-014, D. C. Cook Unit 2, dated Ju'ne 22, 1984.
- 16. PSNH Letter (SBN-1225), dated October 31,,1986, " Response to Request for Additional Information (RAIs)," J. DeVincentis to S. M. Long
- 17. PSNH Letter (NYN-87-002), dated January 20, 1987, " Comments on Draf t
- Report, G. S. Thomas to V. Nerses.
- 18. " VISA--A Computer Code for Predicting the Probability of Reactor Vessel Failure," NUREG/CR-3384, September 1983.
- 19. Residual Heat Removal Experience Review and Safety Analysis, NSAC-52, January 1983.
- 20. NRC Memorandum from B. W. Sheron to Reactor Systems Branch members, " Auto Closure Interlocks for PWR Residual Heat Removal Systems," January 28, 1985.
- 21. NRC Memorandum from W. C. Lyon to C. E. Rossi, " Steam Generator Tube Rupture During Severe Accidents at Seabrook Station," March 3,1987.
- 22. PSNH Letter (SBN-1237), dated November 21, 1986, " Emergency Planning Sensitivity Study," J. DeVincentis to S. M. Long.
- 23. " Evaluation of Severe Accident Risks and Potential for Risk Reduction,"
NUREG/CR-4551, Volume 2, February 1986.
24 NRC Memorandum from B. W. Sheron to D. B. Liaw, " Steam Generator Tube Response during Severe Accidents," February 14, 1985.
i w ~- -
+ -
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2-35
- 25. NRC Memorandum from C. E. Rossi to V. A. Noonan, "Seabrook Emergency Planning - Study--Treatment of Pre-existing Leaks in Containment," Feb-ruary 9,1987.
- 26. Pelto, P. J., et al., " Reliability Analysis of Containment Isolation Systems," NUREG/CR-4220, PNL-5432, June 1985.
- 27. Collins, H. E., et al., " Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Pl ants ," prepared for the U.S. Nuclear Regulatory Conmission, NUREG-0396, December,1978.
- 28. Theofanous, T. G., " Review Comments," University of California, Santa Barbara, dated January 12, 1987.
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', operator recovery of open equipment hatch (calculations performed by PSNH).
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,y PLO-0465 results -
88 g E g 0.01 --
g 3" : \
- g. -
'-l Effects of conservative interpretation by PSNH of He NUREG/CR-4220 data,.PSNH -
C -
letter. .
t I i .
i l
o.coi I ' i' 'i'l i e i i e n iil e i i i i,i, 1 10 100 1,000 OlSTANCE (MILES)
Figure 2.4 Comparison of 200 rem-dose versus distance curves for conservative interpretation by PSNH of flVREG/CR-4220 data. (Calculations performed by PSNH.)
Table 2.1 Summary et Operetsee Events. Emergency care cweing 5tstem. sneletsen meck vegves, teenage Fe:Isre se>de uweer heteresc o et Casch Estleeted ECCS
(*E #3 Plant Dete System valves Leek Rate Event Description neported (gom) pomeras vgl.A.13 Petisodes 5/72 ACC Leoasse Inte si tank. The Interness of a check velve se tne outlet of en 58 teak 1 pS sente 1 mes lacerrectly essemeled.
185.4.25 Mn:a 12/72 ACC Leekage late 58 teak. A smelt piece of sold slag had toeged under the seet of the 1 y5 femte 1 7enkee i
eettet check valve elleelag back leakage. Ottetson: 1700 pre (flelt is 1720 ppe).
] vfl.a.32 Tertey 5/73 wf
! Pelet One et the enroe check volves la to,55 eless devoleped a leekage of t/3 gem. S W 33 Tse other check volves ehemed only slight footage. Faltere of sof t seats.
- 188.A.63 Glene 9/74 ACC Leekage of a check velve caused bores diletten la ACC. "A* Ifrom 2250 ppm to 1617 pas).
1 y20 sk>te 1 I
i Will.A.8$ Surry 1 &/7S
! ACC _ Check velve did not seat. ACC (*lC*) gevel Increased. Leekage rates a4 gas. 1 y10 paste 1 fu VII.A.126 2 Ion 2 10/75 ACC a th ong size goshot teste41ed in the check velve,for ACC. "A*. Leak reto *.2S gps. I leste 1 Ze d 2S o i
Tll.A.105 skingenen 2 1/76 AG: Accumulator (*s") Isleakage through teeklag outlet check velve.
i 1 y20 easte 1 V.A.122 2 ion 1 4/76 ACC lateekepe to ACC. *10* free AC5.
1 y20 Note 1 T l l. A.114 Sorry 1 7/76 ACC Two check volves In serles 81 58 128. 130) Isoked cesslag torce dilution le 2 la serlee y10 testo 1 ACC. *s".
Vll.A.120 Surry 2 e/76 ACC g
Baron difetton (from 1950 ppm to 19933 la 58 ACE. "C" caused by leeking check 2 la series y10 gestes 1
- velves 12-51-145, 147).
and 2 Vll.A.221 Itlllstone 2 4/77 ACC Inleekage et aC througn outlet check volves to Si tank *4* Lou borom 2 In serles y20 esote 2 concentration. Five accurrences an 1977 1 .
fll.A.175 See S/7e LPI TIItIng disk check velve (first velve leside centelament) felled te close alta Deotre 1 gravity testeIIed in a vertical rather thee a horizontal pipeIIne. 1 y3 vlft.A.182 Calvert 9/78 AQ:
j Clifts 2 Outlet check valves for Si tanks 210 and 223 8eemed coroe concentration reduc. 2 y10 esote 1 tion fram 1724 and 1735 ppm to 1652 and 1594 ppm In one month perled. y10
)
=
S
Table 2.1 Continued samner Ref ereac e of Chect Estleeted ECCS Velves Leek Rate SNPE #8 Plant Date System tweet Descristles Reported Igpal Reaerts vil.A.262 Crystal 7/80 ACC Check velve CFV=79 te core fleed tank felled. The $seletten velve to the Ng 1 1004y Notes 1 River 3 system was open for my neulag. a600 gallom ligeld entered the es2 system and c200 and 2
@ gallons ses released. The corresponding activity released estimated es 1.07 eCI.
VII.A.273 Devis 10/90 ACC thel systee isoletles chect volve CF-30 lashed back emessively. Velve It lato. Desse 1 1 S0cyl00 Isote 1 disk and are ked separated from the selve medy. Solts and lachlag mechanten NotIc e were misslag. Core fleed tank overpreneurised.
80-41 Vit.A.291 Surry 2 t/01 ACC Accouletor (*C*) beren diluted. Check velve Il-$f 144) looted. Flushing systee 1 y10 Note 1 leoroperly set up, resultlag la charglag system pressere te eulst se the downstrese slee of the check volve.
Vll.A.303 Pelisados 3/81 ACC Leekage of RC late the $5 teak (T-8233 I yS sestes I fu M2 vll.A.306 De:Culre 1 4/01 ACC w AcCumuletor h a evtlet check velves (ID-l$g and (90-160 were leekleg. RC$ pressures 2 le y10 Note t
, 1800 psig. Acc. pressurer d25 psig water level eteve eiere setpelat. series VII.A.307 stGelre 1 4/Gl ACC $leller evente with Acts. *C" and "D*. 2x2 la y10 footes I series y10 and 2 vi f.A.343 Point 10/81 LPI IICS/LPI feeletica check velve Il-853C) leeks le acess of accepteste criterte 9each I 1 yl0 l>6 geel.
Vll.A 384 Calvert 7/02 ACC Acc. evtlet check velve et Unit I leched due to deterioreflen of the disk seellt.g Cl4tts 1
E200 Note I e-rang. The e-rlag meterial hos base changed en all check velves of Unit I and 2 1&2 1/2 50-219. 229. 231. end 249 v i t. A.40 5 surry 2 g/82 ACC Acc. outlet check volve (2-58-1448 leaked IICS water late tank *C' during a pipe 1 y20 Notes 1 flush resulting la low heroe concentretlee, and 2 Vlf.A.396 Pellsades 9-12/ ACC Minor leekage late $f tank (coopeeded by level Indicottee f ailure) vie check j pS Notes 1 42 velve leekspos.
and 2 1
,Toble 2.1 Coat inued seseher et Deck Estlested Ref erec e ECCS talves Leek Itate luPE #8 Peset Date Systes Event Oescriptlee Reported Egent Demeras vis.A 407 stoutre 1 S/s) ACC RCS =ater leleekege through outlet Check velves lat-170 and in-171. resultlag 2 la series 20s yc50 sente 1 in los borea concentretloa la CLA *G".
til.A.437 Fart 2 9/83 0FI $8 check valve to toep 3 cold leg use emeessively testing, Incomplete contact 1 50sy100 hat to v.t ..s. s t.
LER 64401 Oconee 1 3/84 ACC Accumelstor (*A*) geleonage tareugle leeting volves. Adelaistrative deficiency. 2 in sertos yS esote 1 ne management centrol over a boeum problan (slace S/939.
V.F.0043 Pallsedes 7/04 ACC Accumulator taleshape through feeking check velves CK-3146 and Ot 3114 2 yS Notes I LER 44-012 I
end 2 til.A.452 St. Lucle 12/94 ACC Inleekage to Si tank. Seet plate cocked, volve seet compensating Jelet belt 1 20cycSO sente 1 2 getled, y
a til.A.. C.tv.,,
Cfifis ii. , Ace - 8 0 - e to ...t,ie;ec,ie. t s ,0,,m h clase.
degradeflom lunit 1 = 1.6 gym. Unit 2
- 27.2 gaml.
..e,e.,, ,.ete ... 2 y, 20cy50
-,e i e
1L2 til.A.457 stGuire 1 4/05 A(X: Lee occometeter bores concentratten. I y5 leote I LER 85-007 Pottsedes 6/85 ACC lateatege from the nCS vie e check velve. Lee levet borea concentratten. 1 y5 loote 1 til.A.474 Pollsados 11/85 ACC Accesseletor ($f f-4201 lateskape from Itc$ through a check velve, CK-3814. 1 y5 loote 3 Baron dIIetson.
Note 13 Estimated leek rete is the resultant one through two check volves la serlos.
peote 2 seat IIsted la Tehle 3.8 et PLre0432 esote 3: The Polisados melt has a chrwelc accoulater Inleekage prehlen, t
I l
I 1
l
Table 2.2 bemory of operating Events. Energency Care Cooling System. Isoletten Oneck volves. *Fellere to Itectose teen Demanda Fellere abde weber of Oieck lieference ECCS Velves luPE #1 Pleet Date System Eveet Descrlotten itseerted Aeneres VII.A.270 Segeoyen I 9/00 #9 Si check valve 63-439 ses found to to stock ocea. It ses ceased by 1 smote I laterference between the disk met f eckelre tech sold and the velve body.
V88.A.26S Selen 1 32/30 el $$ caeck velve felled to clase derlag a test, et is se Ieterf ace hetenen RCS I Isote I het leg and $8 peups. Velve ses feend to to leched open det to tarse selldif 4Ca*
.flen declag the test refoollag.
TII.A.294 Oceae,a 2/31 LPs meector vessel LPI leap =0* toeletten valve (IEF-128 leaked escessively during 1 tente I LOCA leek test. The volve elsk had Decess frecen et the plwet la e cached postflat. Belldup of esposit In the gap between the klage and disc haeb consed the freerlag.
VII.A.302 Oconee 3 3/09 LPI SleIIer to event et Italt I (welve levolved is 3 CF-tSt. I saste 3 yll.A.310 secestre 1 S/SI ACC Leek test damaged acc. check velves = seet type changes. 2 Isote i TII.A.318 secGuire i S/03 ACC Acc. check wolves felled. 2 Isote t W WII.A.3fS Pelst 7/el LPS ACS/4.Pl f eeletten check velves 9-OS3 C and D sere feend to be stock la the f ell 2 sente I Spect 3 epee peeltlen. Mlp leakage rete.
Tll.A.392 ABID-2 10/02 sel St Isoletten check velves 2 SS-ISC and 2 S8-12 stock to the open positten derlag 2 ' sente test regnanted by IE testice $1-30. Olsk sted protruded eteve met, disk af seligned, smote la samt listed le Table 3.8 et PLM32.
4
2-44 Table 2.3 Accumulator Check Valve Exposure Data Start of Number of Total Number of Commercial Number of Accumulater Check valve Plant Name operation Years Check Valves (106 Hours)
Arkanseas Nuclear One i December 1974 !!.08 4 3.882 Crystal Elver 3 March 1977 8.83 4 3.094 Davis-Besse i Nevee*>er 1977 8.16 4 2.859 Ocanoe ! July 1973 12.50 4 4.380 Oconee 2 March 1974 11.83 4 4.145 Geenee 3 December 1974 11.04 4 3.882 Rancho Seco April 1975 10.75 4 3.767 Three Mile Island 1 September 1974 11.33 4 3.970 Three Mile Island 2 -
December 1978 7.08 ? 4 2.481 Arkanese Wuclear One 2 March 1980 5.83 8 4.046 Calvert Cliffe i May 1975 10.67 8 7.474 Calvert cittfe 2 Apett 1977 8.75 8 6.132 Port Calhoun September 1973 12.33 8 8.641 Milletone 2 December 1975 10.08 8 7.064 Maine Yankee December 1972 13.08 6 6.475 Felieades December 1971 14.00 8 9.867 St. luete i December 1976 7.08 4 6.363 Beaver Valley 1 April 1977 8.75 6 4.599
- 9. C. Cook i August 1975 10.42 8 7.302
- 9. C. Cook 2 July 1978 7.50 8 5.256 i Fadian Point 2 July 1974 11.50 8 8.059 Indien Feint 3 Au8vst 1976 9.42 8 6.602 Joseph M. Perley I Decesher 1977 8.08 6 4.247 Reveunee June 1974 11.58 4 perth Anna 1 4.058 June 1978 7.58 6 3.984 Prattle letand i December 1973 12.08 4 4.231 Pratete letand 2 December 1974 11.08 4 3.882 Point Beach i December 1970 15.00 4 5.284 Point Beach 2 October 1972 13.25 4 4.643 R. B. Cinna i March 1970
- 15.83 4 5.547 N. B. Behineen 2 March 1971 14.83 6 7.795 Salee ! June 1977 8.50 8 5.957 Surry 1 December 1972 13.08 6 6.875 Sorry 2 May 1973 12.67 6 6.659 Trojan May 1976 9.67 8 Turkey Point 3 6.777 December 1972 13.08 6 6.875 Turkey Point 4 September 1973 12.33 6 6.481 Tankee towe June 1971 14.50 2 2.540 Elen I December 1973 12.08 8 8.466 Ilon 2 September 1974 11.33 8 7.940 McCutre i December 1981 4.08 8 2.859 Sequoyah I July 1981 4.50 10 Sequoyah 2 3.942 June 1982 3.58 10 3.136 TOTAL 2.369(2) l
V 2-45 Table 2.4 Statistical Data on Leakage Events of Check Valves to Accumulators Leak Rate (gpm) Frequency of Frequency of Number of Events Occurrence (per hour) 5 Exceedance 11 10 9 4.64(-7) 1.48(-6) 20 9
3,30(-7) 50 3.80(-7) 1.01(-6) 100 3
1.27(-7) 6.33(-7) 200 1
4 22(-8) 2.53(-7) 2 8.44(-8) 1.27(-7) 3.44(-8) 4 0
9
2-46 Table 2.5 ISL Results Initially Assigned Plant Damage States Frequency Contribution From, Plant Total Damage State VI VS Frequency
-- --LOCA- -- - - - 1. 96 - 4 _ 1.44.-4 . ,
3.4-4 DLOC 1.2-5 0 1.2-5 DILOC 9.8-8 7.7-6 , 7.8-6 ,
8C 2.1-8 0 2.1-8 ^
7D 1.5-7 0 ~ ~ " ' T.57 -
7FPV 7.4-8 1.7-7 2.4-7 1FPV 1.8-8 8.0-7 8.2-7 1FV 8.4-8 5.9-8 1.4-7 Totals 2.1-4 1.5-4 3.6-4 Note:
LOCA: denotes a PDS, which contains those sequences in which the RC leakage in both ISL pathways analyzed exceeds 150 gpm, but does not exceed the RHR system relief valve capacity.
The sequences are essentially medium LOCAs.
DLOC: denotes a PDS, which contains sequences in which the !$L
. is terminated.
DILOCA: denotes a PDS, in which coolant makeup is being supplied to the core, rut the ISL has not been terminated.
The other plant damage states are involving containment bypassing ISLs and core damage.
2-47 Table 2.6 Plant Operational Modes
- Average '
Reactivity % of Rated Coolant Operational Mode Condition, Keff Thermal Power ** Temperature
- 1. POWER OPERATION > 0.99 > 5%
> (TDHR ) F
- 2. STARTUP > 0,99 < 5%
> (TDHR)F
- 3. HOT STANDBY < 0.99 0 ,
> (TDHR)F
- 4. HOT SHUTOOWN < 0.99 0 (TDHR)F>T,,9 >200*F
- 5. COLD SHUTDOWN < 0.99 0 '< 200*F
- 6. REFUELING *** < 0.95 0 < 140*F TDHR = temperature at which the DHR system is initiated (generally 280*F -
350'F
Note many plants do not use standard technical specifications.
- Excluding decay heat.
- fully Fueltensioned in the reactor or withvessel withremoved.
the head the vessel head closure bolts less than k
' 2-48 Table 2.7 Categories of 130 Reported Total DHR System Failures When Required to Operate (Loss of Function) at U.S.
PWRs 1976-1983 No. of Events (% of Events)
Automation closure of suction / 37 (28.5) isolation valves Loss of inventory Inadequate RCS inventory resulting 26 (20.0) in loss of DHR pump suction Loss of RCS inventory through DHR 10 (7.7) system necessitating shutdown of DHR system Component Failures Shutdown or failure of DHR pump 21 (16.2)
Inability to open suction / 8 (6.1) isolation valve Others 28 (21.5)
Total 130 (100.0) 6 9
i
.- 1..
3-1 b
L 3.
. EVAllJATION OF CONTAINMENT BEHAVIOR I'
3.1. Capacity at General Yield Seabrook Containment Building The Seabrook Station containment building (See Fig. 3.1) is a reinforced concrete structure consisting of a basemat, a cylindrical wall and a hemi-spherical - dome.1 The basemat is essentially a 10' thick circular slab which
' supports the cylinder and other internal structures. The cylinder has an internal diameter ,of 140', a height of 149' and a minimum wall thickness of
! 4'-6".
I The dome internal radius is 69'-11 7/8" and the minimum wall thickness is 3'-6 7/8". In pddition, the containment has a mild steel liner on the inside. The liner thickness is 1/4" at the base. 3/8" in the cylindrical
~ '
] wall, and 1/2" in the dome.
j The containment is reinforced with ASTM A615 grade 60 reinforcing bars of
{' various sizes, mainly #18, #14 and #11. The specified yield strength for the reinforcing bars is 60 ksi. Median yield stress and lognormal standard devia-tion obtained from the results of tensile tests are shown in Table 3.1.2 pop the. ultimate strength of #18 reinforcing bars a mean value of 109 ksi and a
-CoV of 0.025 has been reported.2 The liner steel conforms' with ASME SA516 grade 60 for which the specified yield strength is 32 ksi. The mean yield stress was found to be 45.4 kst with a CoV of 0.042.2 At 271 F a mean yield stress of 40.5 ksi and a CoV of 0.065 are reported.2 The mean ultimate i
strength at 271*F was estimated to be 59 ksi and with a CoV of 0.09.2 l'
! The primary membrane reinforcement in the cylindrical wall is divided '
l into two equal groups placed near the inside and outside faces of the contain-ment wall. Each group consists of two layers of hoop bars and one layer of
! meridional bars as shown in Figure 3.2. Since the cylinder basemat intersec-tion is subjected to high bending moments and shear forces, secondary meri-dional reinforcement is placed in this region (See Figure 3.2). In addition, j
two layers of seismic rebars inclined 45' to the vertical axis are placed near
! the outside surface of the cylinder wall. Shear ties are also placed in the
3-2 f
l cylinder near the cylinder-basemat intersection (See Figure 3.2). Major rein-forcement details for the containment wall are sumarized in Table 3.2.
The dome reinforcement follows the cylinder reinforcement until 9.4' above the spring 11ne. Between 9.4* and 79.2* the hoop reinforcement is reduced to one #18 bar near each face. Above 79.2* the hoop reinforcement is terminated and the reinforcement pattern is orthogonal. The meridional cylin-
!' der reinforcement is continued.to 60* above the springline with an increase in its density as the elevation increases. Above 60' every alternate meridional rebar is terminated, and they are bent such that the reinforceinent pattern near the dome apex is orthogonal. Details of the dome reinforcement are shown in Table 3.3.
i Concrete with two design strengths was used in the Seabrook containwent building. A 4,000 psi design strength concrete was used for the basemat, for j the cylinder near the ir.tersection with the basemat, and for both the cylinder i and dome near the dome-cylinder intersection. In the cylinder and dome i
- portions where primary membrane behavior is. expected a 3,000 psi design
! strength concrete was used. ,
I The median and lognormal standard deviation for the 4,000 Tsi and 3,000 psi design strength have been reported for 28-day old cylinders and for aged congrete.2 These quantities are given in Table 3.4 which war. obt.ained from Reference 2. The 28-day median concrete strength values were obtained from
, results of cylinder compression tests for the concrete use6 in the Seabrool structures. The median values for the aged concrete were obtained from the 4
28-day values using the correlation given in Figure 2.1 of Reference 6.
{ Seabrook Containment Model A finite element model of the Seabrook containctent was developed to be used with the computer code NFAP.3 The model is shown in Figure 3.3 and is based on an axisymmetric idealization of the geometry, which is considered a
! good approximation for a structural failure evaluation under axisymetric l pressure loads.
The containment finite element model consists of 408 eight-noded isoparametric elements and 1354 nodes. A set of nonlinear spring i
l i
a -
3-3 elements with a bilinear stress-strain law are used to model reinforcing details such as shear ties. The basemat was considered to be fixed at the bottom nodes.
Throughout the cylinder wall and dome the model has 8 layers of eight-noded elements across the wall thickness, as shown in Figure 3.3. Six layers of elements were used through the thickness of most of the basemat.
The element layers and its properties were chosen to represent separately the liner, the plain concrete, and the reinforced concrete with hoop, meridional and diagonal rebars. Spacing and sizes of the layers have been chesen in order to nodel the actual rebar placements as close as possible. This is particularly pertinent at the cylinder-basemat intersection where high bending moments and shear forces will develop. In addition to these criteria, the modeling requirements commonly used with finite element analysis were also taken into consideration.
The inelastic behavior of the plain concrete is described by tha Chen and Chen elastic-plastic-fracture model.3 Material properties for this model were estimated from the aged (as built) concrete properties and are shown in Table 3.5. Post-cracking behavior of the concrete was modeled using a normal stiff-ness reduction factor a of 10 4, and a shear stiffness reduction factor 8=0.5/ ( c1/ tto) , where ci is the principal normal strain r.ormal to the crack and cto the tensile strain at crack initiation. 8 The shtar stiffness reduction factor is limited to a value not less than 0.10 to ac:ount for the cummulative effect of interface shear transfer and dowel action. The normal stiffness reduction factor a reduces the normal stress diagonal element in the stress-strain matrix, and 0 reduces the shear stress diagonal element in the stress-strain matrix. The tension stiffening effect was modeled with a factor -0.1, which multiplies the concrete Young's modulus.
The elastoplastic behavior of the reinforcing bars and liner steel was modeled by a bilinear stress-strain curve and a Von Mises plasticity model with isotropic hardening. Since the #18 reinforcement bars provide nost of the reinforcement, the mean material properties for these bars were used for all reinforcing bars. For the liner, the mean properties at 271*F were used.
The plain concrete properties used are shown in Table 3.5.
3-4 1.oads included in the analysis are the dead weight of the containment and internal pressure. The dead weight is applied to the containment in the first load step at the beginning of the analysis, while the pressure load is applied to the containment in small increments (5.0 psig) in order to detect the initiation of nonlinear concrete behavior and concrete cracking. Once the concrete cracks its stiffness in the direction normal to the crack plane is reduced by the factor a defined above, and the released stresses are redistributed to the reinforcing steel. As the pressure load is increased the next nonlinearity is the yielding of the liner steel. At this internal pressure the containment is cracked in both the cylinder and dome, and some flexural cracking has been initiated in the cylinder-basemat intersection region.
Based on the results of the analysis described above, yielding of the inside layer of vertical reinforcement at the cylinder-basemat intersection is initiated at 130 psig. At a pressure of 154 psig yleiding of the internal layer of hoop reinforcing is initiated. Above 154 psig the load increment was reduced to 1 psig and yielding in the outside layer of the cylinder hoop reinforcement was observed at a pressure of 157 psig. From 159 psig up to 165 psig internal pressure load increments of 0.5 psig were used. At 165 psig hoop yielding in the cylinder wall extended over a very large portion of the wall in both the inside and outside layers of hoop reinforcement.
At a pressure of 165 psig the strain in the inside layer of vertical reinforcement at the basemat-cylinder intersection has reached 1 percent and the maximum concrete compressive stress at the basemat-cylinder intersection is 5100 psi. The radial shear dowel at the base has reached a strain of approximately 2 percent and the concrete cracking extends beyond the vertical compression reinforcement near the outside face of the cylinder waII. The computer analysis did not result in a numerical instability indicating that higher pressures could still be obtained; however, the pressure increments in the analysis would have to be further decreased. Consequently, it was decided not to continue the analysis any further. The extensive cracking at the base and large strains in the reinforcing indicate that a shear failure at the base is a potential failure mode.
r 35 i
I 3.2 _ Behavior at Large Deformation As discussed in Section 3.1. the containment structure is predicted to reach a general yield state at a pressure of 157 psig, which confirms the
! estimate provided by SMA. As the pressure is increased above this level the containment structure will begin to undergo large deformations. SMA evaluated the behavior of the containment structure at such pressure levels" and the results of this evaluation are summartred in Appendix H.1 of the PSA2 ,
I The hand calculations performed by SMA and used for the probabilistic t
assessment primarily identify several possible weak places in the structure and determine the corresponding maximum pressure capacity in search of the controlling failure mode.
The uncertainties in the results are estimated and l identified as coefficients of variation (CoV) to account for both uncertainty and randomness of material behavior and lack of knowledge regarding the exact structural behavior. The break of the liner plate is defined as the failure l mode. The capacity of the containment structure is computed in terms of the internal accident pressure it can withstand. Any leakage associated with the pressure level is estimated with a CoV.
l Accident scenarios are postulated for both wet and dry containment condi-tions. The corresponding containment liner temperatures are 271*F for the wet l case and 700*F for the dry case. The structural calculations are first per-formed for the wet case and then modified to reflect the reduced material strengths for the dry case. The various failure modes considered in this analysis are discussed and evaluated below. During the course of its review, RNI. observed that the SMA calculations did not show any checker's signature.
As a result PSNH has committed to perform a complete and independent check for all containment strength calculations.
Membrane Failure The cylindrical wall and the hemispherical dome are assumed axisymmetric and they take the pressure load by membrane action. Both the hoop and the meridional pressure capacities are determined based upon the ultimate strength of reinforcing steel bars, failure of which will lead to a gross containment
3-6 failure. The median pressure capacities calculated by SMA at 271*F are as follows:
Mode Pressure CoV cylinder, hoop tension 216psig(governs) .12 -
dome, hoop or meridional 223 psig .12 cylinder, meridional tension 281 psig .12 The govering hoop failure at 700*F corresponds to a median pressure of 198 psig. The above capacities are based on the assumption that the membrane forces are resisted by the reinforcing bars and the liner plate, and not by concrete.
Since the above pressure values correspond to the ultimate strength of reinforcing steel (109 ksi at 7.5% strain), the containment will undergo a great amount of expansion before failure. This is illustrated in Figure 3.4 which plots containment pressure vs. radial displacement of the containment wall as calculated by SMA. At 216 psig the radial displacement of the con-tainment wall away from the base is in excess of 3.0 feet. SMA believes that at this pressure there is a 50 percent chance that the containment liner will remain intact and there will be no gross containment rupture. BNL believes that at these large containment deformations it is difficult to accurately predict the behavior of the containment and that containment liner failure is much more likely. It is also noted from Figure 3.4 that at a pressure of 216 psig the pressure-displacement curve is almost horizontal. Thus, any further pressure capability of the containment would have to be attributed to even greater material strength of the reinforcing steel. Although some reinforcing steel may have a greater strength, BNL believes that for the high strain levels being considered that further consideration must be given to instances of progressive failure of the reinforcing steel.
In the light of the above discussion, BNL considers the 216 psig pressure capacity predicted by SMA to be an upper bound failure pressure. BNL believes that a more suitable median failure pressure should correspond to the pressure level at which the primary membrane reinforcing steel reaches 1 percent strain
3-7 (175 psig for the Seabrook containment). Such a level recognizes the ability of the containment to withstand pressures beyond the general yield, but limits the amount of containment deformation to levels more commensurate with the current state of knowledge concerning containment performance.
The median failure pressure corresponding to the 1 percent strain level for the dry condition is 158 psig as indicated by PSNH in the response to NRC question 20 (PSNH letter dated October 31,1986).
Shear Failure of Wall at Base The shear failure of the cylindrical wall is estimated by SMA at a median pressure value of 319 psig with a CoV of 0.29. This pressure value is deter-mined based upon the yield strength of the reinforcing steel and on the assumption that the critical section will occur at a distance of 0.7 x effec-tive wall thickness above the base. The shear failure corresponding to the ultimate strength of the reinforcing steel is estimated by SMA at a median pressure of 408 psig with a CoV of 0.3.
SMA assigned a large variability to this failure mode due to their uncer-tainty about the applicability of their elastic analysis when some yielding occurs. However, BNL feels that this failure mode is more critical than assumed by SMA. As discussed in Section 3.1, BNL investigated this mods of failure by means of a non-linear finite element analysis. It was confirmed that such a failure is not expected to occur for pressures up to 165 psig.
However, BNL believes that a shear failure at the base is a potential failure mode at pressures above this level.
Flexural and Shear Failure of Rase Slab The flexural capacity of the base slab is determined by SMA based upon the yield line theory. The median basic capacity is estimated to be 168 psig. However, when the friction and mechanical locking between the base slab and the ring girder of the enclosure building are considered, the median overall capacity is estimated by SMA as 400 psig with a CoV of 0.25. Conse-quently, it is concluded that flexural failure of the base slab is not a
_. _ m _ _ _ . ~ ._ . . - - _ _ _ _ - . _ _ , _ _ _ - _.- _ _ _
3-8 4
controlling. failure mode. The' shear strength .o the base slab is also calcu-
{
lated considering restraint from the ring girder. of .the enclosure building.
The median pressure capacity is estimated by 'SMA as 323 psig with a CoV of 0.23. ,
- BNL reviewed the SMA calculations concerning the shear and flexural fail-i ure modes of the base slab and agreed that these failure modes would not be
- controlling.
I Containment Deformations j The deformation of the containment is calculated by SMA based upon the j assumption that concrete will share tensile load with steel even at the ulti-I mate strength of steel. This assumption- presupposes a bond between concrete
! and the reinforcing bars up to the failure pressure except at the location of
! the cracks which are postulated to occur with a spacing of approximately 21 inches. The biaxial tension test results presernted by Julien and Schultzs indicate that concrete will crack at an early steel stress level and the
, deformation at a high steel stress level is due to steel strain only. The ,
{ hoop and the meridional reinforceing bars used in these tests were of same l diameter as those for Seabrook, namely, No. 18 and No. 14. The concrete cracked at a steel stress of 9.4 ksi and the effect of concrete stiffness dis-
' appeared beyond a steel stress of approximately 25 ksi. Consequently, BNL is concerned that SMA may have underestimated the containment deformations corre-sponding to the containment pressure levels. Since the containment deforma-tions can result in containment penetration failures, an underestimation of the deformations would result in higher predicted failure pressures for the
, penetrations. This is so because the failure of penetrations is predicted on j the basis of the amount of deformation they can withstand. The deformation i l capability is then translated into a pressure capability based on the overall i containment pressure-deformation response curve. Thus, if this curve under-
- estimates the containment deformation, the penetration failure pressure asso-ciated with a given containment deformation will be too high.
In response to this concern PSNH provided a comparison of the containment l pressure displacement curve with and without bond stress (RAI 32, PSNH letter :
t,- . -
9 . . -
,r w
dated November 7,1986). This comparison is shown in Fig. 3.4. PSNH con :
cluded that an assumption of no bond stress would have -no effect on the con -
clusions of their studies. The effect..that this assumption . has on the '
reported capability pressures for critical containment penetrations is- dis-cussed in Section 3.3.
3.3 Capability of Penetrations Many penetrations through the containment' shell are provided. These include a few major penetrations such as the . equipment hatch,' personnel airlock and fuel transfer tube and numerous smaller penetrations accommodating system high energy piping, moderate energy piping, electrical, inst'rumentation and ventilation. lines. All penetrations are anchored to sleeves which are embedded in the concrete containment wall. For the major penetrations, the containment wall is thickened into a hub around the penetration sleeve with.
the wall hoop and meridional reinforcing members directed past the opening in p a continuous fashion and additional reinforcement provided as sleeve anchor-age. For each high energy line, the penetration is a forged member, termed a -
flued head, which forms an integral part of the piping and the containment:
sleeve which is welded to the containment liner. For all other penetrations the closure is a flat plate welded to the containment sleeve and either welded or connected with a compression fitting to the penetrating element. These flat plate closures accommodate either single or multiple penetrations.
To assess the capacity of large penetrations, SMA performed an evaluation of the equipment hatch. This hatch is the largest of the large penetrations and was considered to represent the bounding or most critical penetration in this category. In the evaluation, the capacity of the hatch anchorage was determined to be in excess of 300 psi. Possible failure of the liner at the hatch juncture due to sleeve-concrete separation was also evaluated and found to be improbable due to the low magnitude of the predicted liner strain.
These evaluations are considered acceptable.
Although the capacity of the fuel transfer tube anchorage was established in the equipment hatch evaluation, the containment wall in the vicinity of this penetration is subject to punching shear failure since it makes hard
3-10 contact with the fuel transfer building when the containment expands. Using a simple approximation to model the loading and' relying primarily on doweling '
action of the containment reinforcement to resist the load, SMA determined a mean capacity of 320 psig in this failure mode. Acknowledging the approximate nature of this calculation, SMA assigned a large factor of uncertainty to the results. Probabilistic aspects notwithstanding the crude nature of this calculation warranted further verification of the results. Therefore, BNL performed additional calculations for this failure mode to form an independent assessrtent of the important force-displacement parameters.
In the BNL evaluations an approximate model of the system was again used but this mode 1' differed from that used by SMA. ' The results, although differ-ent from those developed by SMA, indicated that no gross deficiencies existed in the SMA calculations. Further, the estimate of the pressure at which contact is made by the containment shell against the fuel transfer building, a controlling parameter in the evaluations, is not subject to the large uncer-tainty associated with the force-displacement parameters mentioned above.
Consequentially the SMA calculations although approximate in nature are con-sidered sufficient to characterize the impact this failure mode has on con-tainment integrity.
To . assess the capacity of small pipe penetrations, SMA performed evalua-tions for three specific penetrations, X-26, X-8 and X-23. X-26 was stated to be a bounding or most critical example of a single pipe moderate energy penetration, X-8 a bounding case for a high energy penetration and X-23 a
- bounding case for a multiple pipe moderate energy penetration. For each case, simplistic inelastic analysis methods were used to estimate the forces developed at the pipe / penetration interface as a function of containment internal pressure. This data coupled with estimates of the penetration failure characteristics allowed the calculation of the probability of penetra-tion failure as a function of containment pressure in each case. The median failure pressure and the associated median leak areas were 166 psig/0. Sin2 ,
180 psig/50in2 and >216 psig/61n2 for X-26, X-8 and X-23 respectively.
The discussions included in the SMA evaluations provided the basis for the SMA contentions that the penetrations evaluated were the bounding cases i
r .__
y
/i .'
3-11 for the fpenetration types considered. Those discussions, however, did not adequately characterize all other penetrations. For this reason SMA was requested to compile a list of all penetrations, categorize them in accordance with design features and demonstrate that the performance of each is adequate-ly represented and bounded by the sample of three evaluated. As a response SMA proiided Table 3.6 characterizing all penetrations and the calculations considered to be pertinent for their qualification. '
A review was made of the evaluations provided for the bounding cases. In each instance the structural aspects of the calculation seemed appropriate, with the exception noted below, but the assignment of leakage area was consid-ered arbitrary. In addition, for each case, failure was induced by the di$ placement of the containment shell. Since the correspondence between this displacement and the containment pressure is dependent on the bonding assump-s , tion made for the containment reinforcement and since BNL has requested SMA to perfowevaluations corresponding to a no bonding assumption (see discussion b in Section' 3.2) BNL elected to further assess the failure pressure and leakage area for thh two penetrations X-8 and X-26. Penetration X-23 was not con-y sidered since it exhibited a high failure pressure.
For thh high energy penetration, X-8, SMA estimated the median failure preisure to be -180 psig for the wet case with 'an associated median leakage area of 50 in2 and a lognormal standard deviation of 0.5.
The estimate of the median leakage area was based on an annular gas of 1/2 S , f:r the full cirewnference, at the containment sleeve. The estimate of the standard devia-tion was arbitrary.
For the no bond case BNL estimates the median failure
' pressure to be 167 psig for the wet case and 152 psig for the dry case. BNL accepts the SMA estimate for the median leakage area but disagrees with the assumption regarding the standard deviation. In the absence of more explicit data concerning the behavior of penetration sleeves at failure, BNL believes that an upper bound for the leakage area approaching the total annulus between
[x the pipe and containment sleeve should be considered. Based on these
[U, considerations failure of this penetration corresponds to a type B failure for the median leak (6 square 19ches to about 0.5 square foot) and a type C failure (greater than 0.5 square' foot) for the upper bound leak.
These s
b i \.
3-12 categorizations agree with the assumptions in the applicant's consequence analysis.
For the moderate energy penetration, X-26, SMA estimated the median fail-ure pressure to be 166 psig for the wet case with an associated median leakage area of'0.5 in2 and -a lognormal standard deviation of 0.69. The estimate of the median leakage area was based on an annular crack of 0.06 in, the machined l clearance between the pipe and the thru hole in the closure plate, extending 1
over 60% of the circumference. The standard deviation was derived by.censid-ering 0.02 inches a miniumum crack width and full circumference cracks. For the no bond case it is estimated that the median failure pressure is 159 psig for the wet case and 147 psig for the dry case. Regarding leakage area, the estimate for the median leakage area is accepted but an assumption for the upper bound leakage area equivalent to that recommended for X-8 should be used. Specifically, consideration should be given to an upper bound for the leakage area approaching the total annulus between the pipe and the contain-ment sleeve. Based on these considerations failure of this penetration corre-sponds to a type A failure for the median leak-(less than 6 square inches) and a type B failure for the upper bound leak (6 square inches to about 0.5 square foot). The applicant's consequence analysis assumed type A for both conditions.
As noted above, one deficiency was noted in the structural evaluations for the penetrations. In those evaluations only a simple concrete shear cone calculation for a generic case was provided to show that the penetration anchorage capacities were adequate. Owing to the highly cracked state of the containment wall at high containment overpressures the reliance on normal concrete action was questioned. SMA was requested to provide additional calculations to demonstrate that small diameter penetration sleeves do not punch through the coatainment wall under the worst pressure conditions assumed in the analysis. The applicant's response to this request was reviewed and independent calculations were performed by BNL to support the conclusion that the anchorage provided for penetrations is adequate.
Another potential failure mode for the piping penetrations, is the fail-ure of the pipe both inside and outside the containment. This failure mode
3-13 was evaluated by SMA for the piping in the sample considered most prone to this failure, the piping passing through penetration X-8. The calculations indicated that the piping failure pressure exceeded the penetration failure pressure. Further given the high ductility of the piping, any- failures of the piping would have gross distortion, crushing and section collapse associated with them limiting the size of the potential leakage areas. These evaluations seem appropriate for the piping ~ considered.
Other piping penetrations involve the containment ventilation and air purge systems and the containment sump system. The containment ventilation lines have isolation valves both inside and outside of the containment. For these penetrations, the most likely mode of failure is considered to be deterioration of the valve sealant materials at elevated temperatures. In the event of the seal failure of the inner containment valve, the volume between the valves must fill and achieve an elevated temperature before failure at the outer isolation valve can occur. The elapsed time for this failure mode is anticipated to be long as compared to other containment failure modes and is therefore considered of little consequence.
The sump system penetration is at elevation -31'6". A review of the drawings originali; provided to BNL indicated that the penetration sleeve is welded to the li Or at the inside of the containment and to the train AaB sump suction valve containment tank on the outside of the containment. As such the sump suction valve containment tank was considered to be a direct extension of the containment vessel and would have to have sufficient capacity to withstand the temperatures and pressures associated with containment overpressurization events. It appeared that SMA did not consider the tank in their evaluations and therefore, they were requested to perform an assessment of the capacity of the sump containment vessel to the accident conditions. In the response PSNH provided drawings that showed that the tank is isolated from the containment atmosphere by a welded plate closure between the penetration sleeve and the suction line piping. Because of its isolation the sump tank is not subject to accident conditions of pressure and temperature and no further evaluation of its capacity is required.
I
3-14 Another type 1 of - penetration is the electrical penetration assemblies-(EPAs). The applicant has indicated that they briefly reviewed these penetra-tions and that they were not a controlling mode of failure. These types - of penetrations -have been included in the ongoing SANDIA test program sponsored by'the NRC.
The NRC staff provided the results of recent tests at Sandia which are summarized as follows:
Test Condition Manufacturer Pressure Temperature Duration Conax 120 psig 700*F 8 days, 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> D.G. OBrien 140 psig 360*F -
9-1/2 days Westinghouse 60 psig 400* 9-1/2 days For' the Conax EPA there was no detectable leakage during or after the test.
The inner seal failed about I hour into the test while the outer seal remained leaktight throughout the test. For the D.G. O'Brien EPA there was no signifi-cant increase in steam leakage during the test period. There was minor and insignificant leakage found during air leak tests following cooldown after the completion of the test. For the Westinghouse EPA there were no detectable
! leaks during the steam pressurized portion of the test or' during the cool-
, down.
4 The results of these tests do not challenge the applicant's conclusion that EPAs would not be a controlling mode of failure. It appears that the
. principal mode of failure for EPAs would be degradation of the seals when exposed to very high temperatures for a long period of time. If such failures occurred, they would be late in the accident sequence and have little conse-
! quence on emergency planning issues.
i i
~' '
3-15 3.4 Summary of Structural Findings .
Based on its nonlinear finite element analysis of the Seabrook 'contain-ment ,8NL concludes that a shear failure at.the. base of the cylindrical wall-is a potential failure mode but would not occur before reaching a pressure of 165 psig. -
BNL agrees that the containment structure would reach a general yield state in the hoop reinforcing steel at a pressure of 157 psig and 'that it is appropriate to consider this pressure as a lower bound pressure for the hoop mode of failure. However, BNL believes that the median hoop failure pressure should . correspond . to the one percent strain level in the hoop reinforcing steel, which is a pressure of 175 psig. The above pressures are for the wet containment conditions. For the dry containment conditions the corresponding median failure pressure is 158 psig and the lower bound pressure (general yield) is estimated to be 145 psig. This latter value _is based on the reduction factor recommendation in Section 11.3.4.1 of PLG-0300.
With regard to containment penetrations, BNL believes that the failure pressures should be based on containment deformations , assuming no bond strength between the. reinforcing steel and concrete. Based on this assumption
~
BNL estimates median failure- pressures for the wet containment condition of 159 psig and 167 psig for . penetrations X-26 and X-8. For penetration X-26, BNL agrees that a Type A (less than 6 square inches) leak path is appropriate for the median estimate; however a Type B (6 square inches to about 0.5 square foot) leak path should be considered as an upper bound estimate. For penetration X-8, BNL agrees that a Type B leak path is appropriate for the median estimate; however, a Type C (greater than 0.5 square foot) should be considered as an upper bound estimate.
For the dry containment conditions, BNL estimated the median failure pressures for penetrations X-26 and X-8 to be 147 psig and 152 psig, respec-tively.
These values are also based on the reduction factor recommended in Section 11.3.4.1 of PLG-0300.
3-16 Although BNL has performed some independent calculations to support its conclusions regarding the containment strength, it also relied on the results of calculations performed by PSNH and its contractors. Therefore, BNL recom-mends that a complete and independent check of all relevant containment strength calculations be performed by PSNH. PSNH comitted to such a check in their letter to the NRC dated October 31, 1986 and has indicated that such a check has been completed.
t 3.5 References .
- 1. Containment Design Report for Public Service Company of New Hampshire.
Seabrook Station Unit Nos.1 & 2, by United Engineering Constructors Inc.,
January 1985.
- 2. Seabrook Station Probabilistic Risk Assessment, Pickard, Lowa and Garrick, Inc., PLG-0300, Appendix H.1, December 1983.
- 3. Sharma, S., Wang, Y. K. and Reich, M., " Ultimate Pressure Capacity of Reinforced and Prestressed Concrete Containments", NU,1CG/CR-4149, May 1985.
4 Hand Calculations by Structural Mechanics Associates (SMA), Originated by RP, dated December 1982.
- 5. Julien, J. T. and Schultz, D. M., Tension Test of Concrete Containment Wall. Elements, Transaction of the 7th International Conference on Structural Mechanics in Reactor Technology, Vol. J, pp. 237-244
- 6. Troxell, G. E., Davis, H. E. and Kelly, J. W., Composition and Properties of Concrete, McGraw-Hill,1968.
e .
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i Figure 3.1 Containment building cross-section.
3-18
- a. 4
- wy,
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- B e CSE (* 5Pa"F ..
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i Figure 3.2 Cylinder reinforcement.
3-19 k
n 1 2345678 !Il i _.
1.!
l~ l l l l f~l'i L.
l I l
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- f I~l, II) ll 1 LINER !
'~
2&5 PLAIN CONCRETE 3&7 HOOP REINFORCE!1ENT _,
4&6 liERIDIONAL REINFORCEltENT 8 DIAG 0flAL SEISti!C REINFORCEMENT [
L b:
a=
l l l l FA,fl I
f i
Figure 3.3 Containment finite element model (NFAP).
250 l I including effeciof bond stress .
(original curve)
^
200 i y / p _
a / - - -
s- d -
assuming no bonding of '
a 150 '
hoop bars I
@ 3 a.
Y l
= i 5
- 100 -) i i T Y o
- B 50 0
1 0 10 20 30 49 Radial Displacement (inches)
Figure 3.4 Pressure-radial displacement relation for containment.
e e
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3-21 Table 3.1 Statistics of Rebar Yield Strength for Various Sizes Bar Size Median Yield Lognormal Standard Stress (ksi) Oeviation
- 4 70.5 .031
- 5 67.9 .031
- 6 67.9 .039
- 7 70.0 .035
- 8 69.6 .034
- 9 70.3 .040
- 10 70.9 .037
- 11 70.6 .040
- 14 73.6 .031
- 18 72.3 .028
- 4 - #11*' 69.8 .040
- Sizes #4 to #11 taken as a group I
I 1
l
\-
l l
l L.
Table 3.2 Reinforcement Details of the Containment Cylinder i
MERIDIONAL !
i i I
Hoop. Primary Secondary seismic Elevation (Both Faces) (Both Faces) (Inside Face) Dpagonal
-30.0' to -15.0' 2-#18 9 12" #18l 0 12"
~
2-#18 9 12" #18 9 11"
-15.0' to -5.0' 2-#18 9 12," #18;9 12" #18 9 12" #k8911" w h
- 5.0' to 80.0' 2-#18 9 12" #18;9 12" ---
A18 9 11" to 119.0' 80.0' 2-#18 9 12" #18 9 12" ---
- 18 9 22" E14922"-
(alternate) i I
I l
O
3-23 Table 3.3 Reinforcement Details of the Containment Dome Hoop Meridional Seismic Elevation (Both Faces) (80th Faces) Diagonal O'(S.L.) 9.4' 2-#18 0 12" #18 0 12" #18 9 22" 9.4' 30' #18 9 12" #18 9 12" #18 9 22"
- 14 9 22" (alternate) 30' 45* #18 0 12" #18 9 10.4" #14 0 19" 45' 60* #18 0 12" #18 0 12" ---
60* 79.2* #18 0 12" #18 9 12" ---
79.2* 90' #18 9 6.4" #18 9 6.4" ---
l-i i
I I
l t
1 i
. Table 3.4 Statistics of Concrete Compressive Strength ,
l ,
l 28-Day Old Cylinders : Aged Concrete
- t i
^
Median Logarithmic
- Median Logarithmic i
Strength Standard Strength Standard Concrete Type (psi) Deviation (psi) Deviation 3000 psi Design Strength Concre,te 4750- 0.14 5700 0.17
- 4000 psi Design Strength Concrete ,
i for Containment 5450 0.10** 6540 - 0.14 i*
na l 4000 psi Design Strength Concrete -
for Tunnels 5780 0.096 6940 0.14 l
l 4000 psi Design Strength Concrete i for Other Structures 5590 ; 0.10 G710 ' 0.14 ;
i j
- Median strength and logarithmic standard deviation are obtained by multiplying the 28 day strength by.a
' random factor, which is assumed to be independent of the 28 day strength and has a median of 1.2 and a lognormal standard deviation of 0.10.
! This number was estimated..
? .
4 '
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- 9 e
3-25 Table 3.5 Concrete Properties MATERIAL PARAMETER f'c=3000 psi f'c=4000 psi Young's Modulus 4340ksi 4650ksi Poission's Ratio 0.19. 0.19 Yield Strength in Uniaxial Tension 0.233ksi 0.262ksi Yield Strength in Uniaxial Compression ~ 2.46ksi 2.81ksi Yield Strength in Biaxial Compression 2.85ksi 3.27ksi Fracture Strength in Uniaxial Tension 0.54ksi 0.61ksi Fracture Strength in Uniaxial Compresston 5.7ksi 6.54ksi Fracture Strain in Tension 0.00045 0.0005 Frteture Strain in Compression 0.005 0.005 1
i Table 3.6 Characterization of Containment Penetrations Penetration Closure Penetration Number Penetration Specifically Qualification Type Analyzed Method I. Flued llead X- 1 t o X-8, X-9 t o X- 1-5, X-8 Report pages H.1-44 to X-63 to X-66 (18 inch, sch 100 -Carbon Stee'l) ' N.1-50 II. Flat PlateClosure I-25, X-26, X-27 X-26 Pages H.1-39 to H.l-44 Thick Wall - Large (4 inch, sch 160, stainless)
Bore Piping ,
III. Flat PlateClosure X-16 thru X-24 X-23 Pages H.1-39 to II.1-43 oa Th i n Wa l l - La rge X-28 thru X-34 (12 inch, sch 40, Carbon Steel) da Bore and Small Bore X-39, 41, 42, 50, 60, l Piping 61, 67 ;
I IV. Flat Plate Closure X-35 thru X-38 X-71 k Page H.1-37, H.1-39 Thin Wall Piping X-4 0, X-4 3, X-47, X-48. -
Multiple Penetration X-49, X-50, X-52, X-57,' I X-71 thru X-76 l t
V. Fuel Transfer Tube X-62 X-62 l Page H.1-50 to H.1-55 1
9
41
- 4. CONTAINMENT. EVENT TREE REVIEW In this section the ability of the Seabrook containment to contain severe accident loads is examined. Note that ways in which the containment might be bypassed or not isolated are discussed separately in Section 2. This section, therefore, specifically' deals with ways in which' severe accident loads might result in loss of integrity ' of the Seabrook containment. The section 'is divided.into two major parts. Firstly, potential containment loads are iden-
'tified and discussed. Then, the applicability of these loads for Seabrook is briefly summarized.
The bottom line of the updated assessment of containment performance for Seabrook is given in Table 4.1, which 'was reproduced from PLG-0465.1 The con-ditional probability of.a gross early containment failure given a core mel't accident was predicted to be 0.001 at Seabrook compared with 0.34 in the RSS.2 The probability conditional on core melt of early failure in PLG-0465 is an order of magnitude lower than in the SSPSA.3 This is largely due to the reduced frequency. (relative'to the value. in the SSPSA) of interfacing system LOCAs, which are discussed separately in Section 2. '
~
Note that the containment event tree quentification 1n the ' SSPSA3 was
~
reviewed at BNL in NUREG/CR-4540. The BNL review was limited in scope and did not include at that time a detailed assessment of the Seabrook containment !
behavior. This has subsequently been performed as part of the present review, and it is documented in Section 3. However, the review of the SSPSA was
)
sufficiently detailed to ' allow BNL to conclude:
"There is negligible probability of prompt containment failure. Failure during the first few hours after core melt is also unlikely and the timing of ;
overpressure failure is very long compared to the RSS. Most core melt acci-h dents would be effectively mitigated by containment spray operation."
h q
The. above conclusions were not based on Seabrook specific calculations performed at BNL but reflected our best judgment based on extensive reviews of other similar containment designs (in particular, our review 5 of the Zion Probabilistic Safety Study8 ). In this section, we critically review the above i
4-2 I
L conclusions based on our current understanding of containment loads and performance during severe accidents.
{
t 4.1 . Potential Containment Loads During a core melt accident,.there are several possible types of contain-ment loads that could occur. Each are briefly discussed below:-
H, combustion: During a core melt accident, significant quantities of H2 g and other combustible gases could be generated. If these combustible gases - l accumulated to large concentrations before igniting, the resulting deflagra-tion could impose a high pressure / temperature load on the containment. The applicant presented information to indicate that such loads would not serious-ly challenge the Seabrook containment. This potential threat to containment ,
integrity is discussed in more detail later in this section.
Steam /noncondensible gas partial pressures: Without the containment heat removal systems operating, steam and noncondensible gases generated during the core melt accident would cause the pressure in containment to increase. At the time of reactor vessel failure, there is the potential for the hot core debris to contact water. This contact could result in rapid boiling of the water and a sharp pressure pulse. in containment. Limiting calculations were performed by the NRC sponsored Containment Loads Working Group (CLWG), which 7
demonstrated that the pressure pulses resulting from quenching the core debris by boiling water would not pose a threat to the Zion containment. BNL considers- these calculations to be applicable to Seabrook and, as the calculated peak load is much lower than the general yield of the Seabrook containment (i.e., 157 psig, refer to Section 3), we conclude that this containment load is also not a threat to the Seabrook containment.
Steam explosions: When molten core materials fall into water, experi-ments indicate that the boiling can become explosive in nature. It has been postulated that these explosions could generate missiles which could directly fail the containment boundary. The potential for an invessel steam explosion to occur and generate a missile capable of failing a containment building was investigated by a group of experts, and the results were published in
4-3 NUREG-1116.8 The conclusion of this expert group was that such events have a relatively' low probability. The results of this expert group are. consistent with the applicant's submittals on Seabrook.- The allocation of a very low probability (10 " conditional on core melt)'to this event was considered to be reasonable by the BNL review team.
Direct containment heating: This is an area of significant phenomenolog-ical uncertainty related specifically to core meltdown with the primary system
, at high pressure. It has been suggested 8 that if core materials are ejected-from the reactor vessel under pressure .that they form fine aerosols, which could be dispersed into the containment atmosphere and directly heat it. An additional concern is the oxidation of the metallic content of the core debris. These reactions are very exothermic and would add an additional heat load to the containment. The pressure rise.in containment due to direct heat-ing is proportional to the quantity of core debris ejected from the reactor vessel which is finely dispersed into the containment atmosphere. The appli-cant considered that this phenomena is not a concern at Seabrook. because of the design . configuration of the containment, which they felt would prevent dispersal of the core materials into the bulk of the containment atmosphere.
The applicability of this phenomena to the Seabrook containment is discussed
) in the following section.
4.2 Application to Seabrook The combined probability (conditional on core melt) of the above phenome-na resulting in.early containment failure was determined by the applicant to be less.than 10 " for Seabrook. In order to check the validity of the appli-cant's estimate of early containment failure, it would have been necessary for BNL to . develop Seabrook-specific containment loads and combine them with our assessment of the structural capacity of the Seabrook containment. :However, given the scope of the present review, RNL has not developed Seabrook-specific containment loads. Thus we were not in a position to validate the applicant's estimate of early containment failure.
However, BNL has been involved in updating (NUREG/CR-4551, Volume 5) the risk profile for the Zion plant for input to the NRC's " Reactor Risk Reference
, - + .%. , - . - - - - - . . -, -n . - - , - , , - - - , , - - - - - , , - - -. - - , . -, , - . - -
4-4 Document," NUREG-1150.10 The updating of risk for Zion was based on a method-ology developed as part of the Severe Accident Risk Reduction Program I (NUREG/CR-4551, Volumes 1-4) at Sandia National Laboratory (SNL). This i methodology' used expert judgment in an attempt to estimate the uncertainty associated with determining containment loads. The methodology was developed at SNL specifically for the Surry plant but was extrapolated to Zion at BNL The Zion plant. is very !iimilar to Seabrook in terms' of the containment volume i to reactor power ratio. Thus, extrapolation of the Zion loads to Seabrook
- would give some indication of the impact of applying this new methodology to l
Seabrook. It must be emphasized that this exercise should in no way be
, interpreted as a Seabrook-specific calculation. It simply gives some indica-tion of the sensitivity of the Seabrook results to the types of uncertainty in i estimating containment loads discussed in NUREG-1150. It should also be noted that this work is preliminary and has not yet undergone full peer review outside of. NRC and its contractors. It is, th'erefore, subject to revision.
t The range of containment loads reported in Volume 5 of NUREG/CR-4551 for Zion is very wide and far exceeds the loads that would be considered credible by the applicant for Seabrook. Of particular interest is the loads at the
- time of reactor pressure vessel failure. These loads can range from about 60
. psia to 200 psia depending on whether core melt is occurring with the primary
! system at high or low pressure and on whether or not containment heat removal systems, CHRS (sprays and fan coolers) are operating. The higher containment f loads are postulated to occur for accidents in which the primary system pres-i sure remains high immediately before reactor pressure vessel failure. For these accidents, direct heating of the containment atmosphere by core debris or, hydrogen combustion with a steam spike at the time of reactor vessel fail-l ure are possible mechanisms for failing the containment. The applicant has presented information which indicates that these mechanisms are not credible
! ways of failing the Seabrook containment. However, as noted above, BNL does not have Seabrook-specific containment loads so we cannot, at this time, elim-inate these mechanisms as potential ways of failing the Seabrook containment.
For accidents with the primary system at high pressure and without the CHRS operating an approximate median load of 135 psia (120 psig) was predicted j for Zion.- If this median load is compared against the capacity of the 1
~ _ - - - . . - _ - -. ---. - . - -
4-5 Seabrook . containmenti given by the BNL review, one would conclude that the l potential for:early containment failure at Seabrook is very low and would not
. influence. the risk estimate in PLG-0465. However, the range of loads esti-mated for Zion implies considerable uncertainty. The 95th percentile estimate of the probability of early containment failure at Zion ~ is quoted as 0.17 in Volume 5 of NUREG/CR-4551. If this early containment failure probability were also true for Seabrook, the risk estimates in PLG-0465 would increase 'signifi-cantly. However, the capacity of the Seabrook containment ' is greater than Zion (the general yield for Seabrook is 157 psig compared with 134 psig for
~
Zion) so the 95th percentile estimate of early containment failure should be
, lower at Seabrook than Zion. However, BNL cannot at this time quantify how much lower because we have not quantified Seabrook-specific containment event trees -.with Seabrook-specific containment loads combined with our estimate of the structural capability of the Seabrook containment, 4.3 Sumary BNL did not develop Seabrook-specific containment loads given the scope of the current review. However, BNL did develop Zion-specific containment 4
loads as part of updating (NUREG/CR-4551, Volume 5) the Zion risk profile for input to NUREG-1150. As the Zion plant is similar to Seabrook, it was decided to use the Zion-specific loads to give some indication of the sensitivity of the Seabrook containment to the types of uncertainty in estimating containment
- loads identified in NUREG-1150. The range of loads reported in NUREG/CR-4551
{ is very wide (60-200 psia). However, if the median Zion load is compared with
[ the capacity of the Seabrook containment given by the BNL review, the poten-tial for early containment failure at Seabrook is predicted to be very low.
However, the range of Zion loads implies considerable uncertainty in esti-mating the probability of early containment failure. Most of this uncertainty is given by accidents in which the primary system pressure remains high imme-I diately before vessel breach. For these accidents direct heating of the con-tainment atmosphere by the core debris or hydrogen combustion with a steam i
spike at the time of reactor vessel failure have been postulated as mechanisms which could fail the containment. The applicant has presented information which indicates that these mechanisms are not credible ways of failing the Seabrook containment. However, as BNL has not developed Seabrook-specific i
- - , , ,, , . , - - - . , - - - - - - - _ - . , , , , , , . - - - - - ,--,-..-------n~ - = .n - . , - - ,. , , - - - - -
4-6 containment loads, we cannot confirm that the uncertainty associated with pre-dicting the probability of early containment failure at Seabrook is as low as that claimed by the applicant.
4.4 References
- 1. "Seabrook Station Emergency Planning Sensitivity," PLG-0465, April 1986.
- 2. U.S. Nuclear Regulatory Commission, " Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear.. Power Plants," WASH-1400, NUREG-75/014, October 1975.
- 3. "Seabrook Station Probabilistic Safety Assessment," Pickard, Lowe and Garrick, Inc., PLG-0300, December 1983.
4 Khatib-Rahbar, M., et al., "A Review of the Seabrook Station Probabilis-tic Safety Assessment: Containment Failure Modes and Radiological Source Terms," NUREG/CR-4540, RNL/NUREG-51961. February 1986.
- 5. " Review and Evaluation of the Zion Probabilistic Safety Study." NUREG/CR-3300, Volume 2, July 1983.
- 6. " Zion Probabilistic Safety Study," Commonwealth Edison Company, September 1981.
- 7. " Estimates of Early Containment Loads from Core Melt Accidents," NUREG-1079, Draft Report for Comment, December 1985.
- 8. "A Review of the Current Understanding of the Potential for Containment Failure from In-Vessel Steam Explosions," NUREG-1116, June 1985.
- 9. " Evaluation of Severe Accident Risks and Potential for Risk Reduction,"
NUREG/CR-4551, Volume 5, February 1986.
- 10. " Reactor Risk Reference Document," NUREG-1150, Draft for Comment.
February 1987.
- 11. " Evaluation of Severe Accident Risks and Potential for Risk Reduction,"
NUREG/CR-4551, Volumes 1-4, February 1986,
- 12. PSNH Letter (SBN-1237), dated November 21, 1986, " Emergency Planning Sensitivity Study," J. DeVincentis to S. M. Long.
5
_. _ _ , . --. - - - - . . . - _ ._._ ,..--m-_ . . . - _ _ _ -
4-7 Table 4.1 Comparison of Core Melt Frequencies and Distribution of Release Types (reproduced from Table 2.2 of PLG-0465)
Risk Parameter '. A -
00 Updated 33p34 PWR Results*
e Mean Core Melt Frequency (events 9.9-5** 2.3-4 2.7-4 per reactor-year) e Percent Contribution of Release Types -
Gross, Early Containment 34 1 0.1 Failure Gradual Containment 66 73 60 Overpressur'ization or Melt-Through Containment Intact 0 26 40
- 8ased on RMEPS (PLG-0432).
- Based on WASH-1400 uncertainty ranges.
NOTE: Exponential notation is indicated in abbreviated form; i.e., 9.9-5 = 9.9 x 10-5, 4
[ 5-1
- 5. REVIEW 0F SOURCE TERMS 4 In this section the fission product source terms developed for PLG-0465 1 are reviewed. The source terms used in PLG-0465 1 are reproduced in Table
} 5.1.- The probabilities of each of the source terms.are given in Table 5.2.
5.1 Fidelity to WASH-1400 Methodology f
The fission product release fractions in Table.5.1 were determined by the applicant using RSS2methodology. These source terms are consistent with the point-estimate source terms used in the SSPSA.3 The SSPSA source terms were reviewed by SNL in NUREG/CR-4540" and they were found to be reasonable given the limitations of the RSS methodology (principally the CORRAL5 code).
i I
I When raviewing the PLG-0465 source terms, questions were raised and L
! transmitted,to the applicant. One question related to release category S2W (refer to Table 5.1). The fractional release of the noble' gases for S2W was quoted as 0.123 whereas the release fractions for Cs and Te were quoted as 0.2 l
and 0.19 respectively. It appeared inconsistent to release more aerosols (Cs j and Te) than noble gas, and the applicant was requested to either explain the l predictions or provide revised source terms. In reference 6, the applicant i provided a response to this question. Basically, the noble gas release in p Table 4.3 (reproduced in Table 5.1 of this report) of PLG-0465 was a mis-3 print. The noble gas release fraction for the S2W-3 release phase should have ,
j been 0.23 rather than 0.023. However, as the S2W-3 release phase occurs very-late (approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />) after reactor scram, the impact on risk of this l larger noble gas release would be very small. Calculations performed at BNL I have verified that the impact of increasing the noble gas release in the S2W-3 phase on both the dose versus distance curves and the risk profiles is negli-
. gible.
l S.2 Credit for Scrubbing of Submerged Releases Another question related to the credit assumed for the interfacing system i LOCA source term in which the break location was assumed to be submerged under l water (S7W in Table 5.1). In PLG-0465, the source term mitigation resulting l i
4
_ ~ . , _ _ _ _ _ _ _ _ . - _ _ _ . , _ . _ _ _ , _ - _ _ _ _.-_-. _ _ _ ___._ _. -
s t 5-2 l
from a subcooled 30 foot deep pool was modeled as a decontamination factor (DF) of 1000 for all release fractions except the noble gases. In the RSS, a decontamination factor of 100 was used for fission product scrubbing in a sub-N cooled pool. Thus in order to be consistent with the RSS methodology, it l
appeared that a lower DF should be used. However, if the pool were indeed subcooled, calculations at BNL indicated that using a DF of 100 rather than '
1000 would have no impact on the risk calculations presented in PLG-0465.
! A more important question was whether or not the pool would.be subcooled
- or saturated. In the RSS, no credit was given for fission product scrubbing
! in a saturated pool, and therefore, the applicant was requested to provide justification for the subcooled assumption. In reference 7, the applicant
. provided a response to this question. Arguments were provided to indicate that the pool would be at least 10*C subcooled and that this degree of sub-cooling together with the large pool depth was sufficient to justify a DF of 1000 rather than a DF of 100 was used in the, RSS. However, the objective of the question was primarily to determine if the pool would be subcooled and i based on the response, this appears to be the case.. BNL had already concluded
- that changing the DF from 1000 to 100 would not change the dose versus distance nor the risk profiles. Finally, the conclusion . (given in reference j 7) that even if pool decontamination were completely ignored, that the dose
- versus d.istance and the risk profiles in PLG-0465 would not be significantly effected was examined at. BNL. Calculations at BNL indicated that if the fre-
{ quency of interfacing systep LOCAs reported in reference 1 was used, then the l
conclusion was correct. However, if the revised -frequencies for interfacing systems LOCA suggested in Section 2 by the BNL review were used, then com- l plately ignoring pool decontamination would impact the risk estimates.
i l 1
5.3 Uncertainties 2
The applicant considers the WASH-1400 source terms used in PLG-0465 to be very conservative and the applicant has high confidence that the source terms l would not be exceeded in a real accident. BNL found the source terms used in
! PLG-0465 to be consistent with WASH-1400 methodology but we are not as confi-
- dent as the applicant that they could not be exceeded. The new source term methods (refer to NUREG/CR-4551,8 Volumes 1-5) indicate that if the i
e i .____ - - -
- e 5-3 containment fails late. or if there is gradual leakage from containment then the aerosol fission product release is likely to be lower than would be pre-dicted by WASH-1400 methods. This is because WASH-1400 methods underpredicted aerosol . agglomeration and settling. Therefore, if the new methods were applied to release categories S2W and S6W, the predicted aerosol release would be lower than WASH-1400 values. However, the new methods also indicate- that if containment fails early and the opening' is large, then there is consider-able uncertainty associated with predicting fission product release. The uncertainty ranges associated with fission product release in NUREG/CR-4551 can, for certain acc.ident sequences and early containment failure modes, exceed the WASH-1400 predictions. This uncertainty would principally affect the S1W release category at Seabrook.
5.4 Summary 1
In summary, the fission product source terms used in PLG-0465 appear in general to be consistent with the approaches used in the RSS. The misprint in Table 4.3 of PLG-0465 related to the fraction of noble gas release was found by the applicant (and confirmed by BNL) to have negligible impact on the risk profiles or the dose versus distance profiles reported in PLG-0465. Thus, the corrected noble gas-release fraction in the S2W release category would, by itself, not change the conclusions in PLG-0465. In addition, the argument d
presented by the applicant that the water in the RHR vault is sufficiently subcooled to warrant consideration of significant decontamination appears reasonable. Finally, the statement in the applicant's response that even if pool decontamination had been ignored, the risk profiles or dose versus distance profiles would not change significantly was confirmed at BNL. Note, however, that this conclusion is based on the frequency estimates for inter-facing system LOCAs in PLG-0465.
5.5 References
- 1. "Seabrook Station Emergency Planning Sensitivity Study, PLG-0465, April 1986.
1
e <
5-4
- 2. U.S. Nuclear Regulatory Commission, " Reactor Safety Study: An Assessment of ' Accident Risks in U.S. Commercial Nuclear Power . Plants," WASH-1400, NUREG-75/014, October 1975.
- 3. "Seabrook Station Probabilistic Safety Assessment," PLG-0300, December 1983.
- 4. Khatib-Rahbar, M., et al., "A Review of _ the Seabrook Station Probabilistic Safety Assessment: Containment Failure . Modes and Radiological Source Terms," NUREG/CR-4540, February 1986.
- 5. Burian, R. J., and Cybulskis, P., " CORRAL 2 User's Manual," Battelle -
Columbus Laboratories, January 1977.
- 6. PSNH Letter (SBN-1234), dated November .17,1986, " Response to Request for Additional Information (RAIs)," J. DeVincentis to S. M. Long.
l 7. PSNH Letter (SBN-1227), dated November 7,1986, " Response to Request for Additional Information (RAIs)," J. DeVincentis to S. M. Long.
i l
l i
t
) >>
^
Table 5.1 Release Categories for Seabrook Station Based on WASH-1400 Source Term Methodology'
,,3,,,, Release Release Warning Ener9y.
Time Release Fractions Duration Time Release
- '9 ##
, (hours) (hours)- (hours) (MCA/S) XE 0.I. 1-2 CS TE 8A RU LA i
j SIW 2.5 0.5 1.0 11.9 _ 0.9 7-3 .7 .5 .3 .06 .02 4-3
$2W-1 4.8 2.0 0.5 0 .03 2.1-4 4.3-3 l 52W-2 6.8 4.0 2.5 0
.023 4.2-3 2.8-3 8.4-4 8.4-5
.07 5.0-4 1.3-3 .048 .039 5.5-3 3.4-3 52W-3 19.8 18.0 15.5 0 .023 1.6-3 5.2-4 2.3-3 .126 .147 .014 .011 1.9-3 1
TOTAL 4.8 24.0 0.5 0 .123 2.3-3 7.9-3 .20 .19 .022 .01 5 2.5-3 1
', 53W 6.0 24.0 2.0 0 4.7-4 3.3 3.2-5
- 1.7-4 1.5-4 1.9-5 1.2-5 2.0-6 ,
, S6W-1 1.75 1.0 1.5 0 .15 1.1-3 i 56W-2 2.75 4.0 2.5 .10 .11 02 .014 4.1-3 4.1 -4 0 .42 2.9-3 .07. .19 .063 .022 56W-3 15.75 18.5 15.5 0 .009 .001 m
.32~ 2.2-3 .01 .13 .32 .011 .020 .3.8-3 '
+
TOTAL 1.75 23.5 1.5 0 .9 6.2-3 .18 .43 40 .047 .033 5.2-3 57W 8.5 7.0 2.0 0 .9 7-6 7-4 5-4 3-4 6-5 2-5 4-6 1
j NOTE: Exponential notation is indicated in abbreviated form; f.e. 7-3 = 7 x 10-3 _
i V
l l
- 3. g . ~ .
3 . -
yys
.y \
n , i
_g ,
e
- ,y Table 5.2 Revised C-Matrix fdr New Source Term Categories
~
- n , c i ~ t-
,'.4Y,
" .7 Plant OW '
Nllfi Damage- .
Source Term Category
% State '
+'
S1 S2 S3 o
'SS N (frequency) c 56 57 M3M n.
IF l' ( j 1.0 (f..!,9 1
( 2.0-8)
' ~
(2.0-8) i t t
{if 1 ipy. : i,g .
"1 h (4.6-9) -(4.6-9) 1FP 1.0
-(1.4-6) (1.4-6) i IFPV 1.0
,j ' ( 2.7 -8 )
4 (2.7-8) 2A . 3.4-5 1.4-4 '1.0-2 0.99
-(1.9-6) (6.5-11) (2.7-10) (1.9 8) (1.9-6).
.30/70 2.0-6 8.0-5 0.95 0.05
'(J.8-5) . (7.6-11) (3.0-9) ( 3.6-S ) s(1.9-6)
?
3F/7F 1.0
, ( 3.0-7 )
(3.0 7) ti lO 3FP/7FP' 1.0 3' (1.9-5) (1.9-5)
W 4A/8A 3.1-6 1.3-4' 5.2-3 0.995 m'
,9 , (1.1-4) (3.3-10). (1.4-8) (5.5-7)- (1.1-4) 9 7FPV 1,0
.;, ~ ( 1.2-3 ).
(1.2-8) 80 1.1 '6 3.1-5 0.9999
( 1.0-4 ) (1.1-10) ( 3.2-9 ) (1.0-4)
- g Total 5.2-9 2.0-5 1.4-4 1.1-4 3.2-7 FreqLency 3.9-8 7
i
- 'N0TES:
r,,
- 1. Exponential notadion is ind!cated in abbreviated form; i.e., 2.0-8 = 2.0 x 10-8
' 2. Numbers inside parentneses are unconditional frequencies (events per reactor year) bas'ed on mean values. Numbers not inside parentheses are conditional by
'~ frequencies of source term categories, given the indicated plant damage state, also c in,Section based3. on mean values. Median values of source term categories are presented i r-
- x j ,
3 -e g_
%! ,,:l ' -
$ i l- 6-1
- 6. S{TE CONSEQUENCE MODEL 6.1 NUREG-0396 Basis ,
NUREG-0396 1 introduced the concept of generic Emergency Planning Zones (EPZ)' as a basis for the planning of response actions which would result in dose savings in the immediate vicinity of nuclear facilities in' the event of a serious power reactor accident. The actions would be triggered if projected
!' radiation doses to an individual would exceed Protective Action Guides (PAGs),
as ' discussed and referenced in -NUREG-0396, although ad hoc actions could be taken at any time. The PAGs are 1 to .5 rem whole body dose and 5 'to 25 rem thyroid dose hut are not intended to represent acceptable dose levels. Fur-thermore, protective actions may not assure that PAG levels can be prevented.
,It was concluded in NUREG-0396 that the objective of emergency response plans should be to provide dose savings,for a spectrum of accidents that could
-produce offsite doses in excess of the PAGs since no specific accident could be identified as the one for which to plan.
B 4
' The most important guidance for emergency planners is . the size of the EPZ.
Based on factors that included risk, probability, and accident conse-quences, it ,was . judged that a generic distance of about 10 miles was appropri-ate' for core melt accidents. The less severe accidents would not have conse-c quen' es in excess of the PAGs beyond this distance, whereas the more severe accidents would not in genera 1 cause early injuries or deaths beyond this ~ dis-tance. Hence, protective actions were judged to .be most useful within this distance.
NUREG-0396 used the accident release categories of the Reactor Safety Study 2 (RSS) to compute the risk of exceeding various dose levels in the absence of protectiv.e actions for a spectrum of accidents. The RSS accidents and their median probabilities are given in Table 6.1. Using the original RSS consequence model ('CRkC)3 and the accidents PWR1 through PWR7, a 200 rem whole body' risk curve was -constructed, shown as the heavy line marked 0396 in Figure 6.1. This is the level at which serious injuries and sone deaths can occur.
9 4 6-2 j J
l However, CRAC did not compute the 200 rem risk directly. The authors of
. NUREG-0396 had to interpolate to obtain the 200 rem curves. Interpolation was performed for each component accident to obtain the conditional risk given the f.ccident. Then the conditional risk at various distances was multiplied by the probability shown in Table 6.1 and divided by the core melt probability of l 5x10 8.per reactor year. The results for each were qummed to give the overall risk of the accident spectrum. Each risk was computed using about 100 weather samples from a typical series of New York City. hourly annual weather data with the assumption that people would follow normal activities for one day follow-
- ing arrival of the first plume to reach their location. That is, people would stay at their original location and receive groundshine doses for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
BNL recomputed the 200 rem dose vs distance curve using an updated ver-sion of CRAC, called CRAC2," and a more detailed grid. The conditional compo-nent risks were computed directly (no interpolation required) and the results for each of the WASH-1400 release categories are shown in Figure 6.1. PWR6 and PWR7 did not exceed 200' rem.
The overall 200 rem risk curve was then computed using a core melt proba-bility of 6x10 5 per reactor year, which is the sum of the probabilities for PWR1 - PWR7 given in Table 6.1. The curve gives higher risk for 1-3 miles and lower ri.sk for 4-10 miles than NUREG-0396, but the curve still drops sharply
- beyond 'about 10 miles. It should be noted that most of the core melt proba-l bility comes from PWR7 which does not contribute to the 200 rem curve. It can be concluded that the approximations used in NUREG-0396 are not substantially
- different from the more detailed calculations done by BNI. using CRAC2.
'62 Consequence Modeling l
The applicant used the CRACIT6 code for their consequence assessments in
[
PLG-0465.10 In this section CRAQIT predictions are compared against conse-quences models currently being used by the NRC and their contractors, namely
- CRAC, CRAC2, and MACCS.6 The factors involved in consequence modeling are discussed in Appendix 6 of the kSS. All codes compute early and delayed health effects from cloudshine, inhalation, and groundshine. The early health effects are based on data from the Marshall Islands, bomb tests, clinical data l
- -_ - . .- .. _- .. - - ._ ~.
- 3. o 6-3
. from radiation therapy, and lab animals (particularly for lung data). The.
.three CRAC models (CRACIT, CRAC2, and CRAC) use a stepwise linear function with a threshold dose for early effects as discussed in the RSS. The MACCS code (recently developed at Sandia National Lab) uses . a hazard function approach without a threshold at discussed ip'NUREG/CR-4214, The latent
. effects in the CRAC'models are calculated from the BEIR-1 reports which uses Japanese data plus a modification to the linear dose response curve to account 8
for reduced effectiveness at low doses. LMACCS uses' the BEIR-3 model which is a linear quadratic dose response model with absolute risk of cancer for some organs such as bone marrow and relative risk for other organs, depending on
, population makeup. In addition, CRAC and CRAC2 allow 'only a one " puff" release of radioactivity whereas CRACIT. and MACCS allow "multipuff" releases.
There are also other differences in the codes, such as the shape of the plume,
, dry and wet deposition of particulates.- weather sampling, resuspension, etc.
which can account for differen'ces of a -factor of two in the results.
BNL used the MACCS code and the source terms defined in Table 5.1 to review the calculations presented by Seabrook in PLG-0465 10 using_CRACIT. The
' comparisons are based on the 200 rem dose probability vs distance curves using the' source terms and weather data supplied by Seabrook. In addition, BNL cal-culated the individual risk of exceeding 5, 25, and 300 rem to the thyroid and 5
the individual risk of death as a function of distance. In all cases, it was assumed that the population was exposed to one day of groundshine following
- arrival of the first plume segment. They would also be exposed to other L
I plumes that. arrived at their location within the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- j. 6.2.1 Whole Body Dose Vs Distance j:
The MACCS code does not have an organ defined as "whole body" so red i marrow was used as a substitute. In CRAC2 calculations it was found that the
! red marrow dose was about 30% higher than the whole body dose. Also, early I health effects are sensitive to the red marrow dose. Thus, the red marrow l dose'is a good substitute.
MACCS does not directly calculate the risk of a dose vs distance, and it was necessary to define an appropriate risk function to obtain this curve. In
6-4 the MACCS calculations, the risk of exceeding 200 rem to the red marrow was set to zero if a weather sample yields a mean dose less than 200 rem and one if a weather sample yields a mean dose greater than 200 rem. Thus the mean risk is dependent upon the annual weather data. The risks were calculated for the Seabrook release categories S1W (one puff), S2W (3 puffs), and S6W (3 puffs). The results are given in Figures 6.2, 6.3, and 6.4. The conditional probability of .001 risks extend to 25 miles for S1W, to 2 miles for S2W, and to 4 miles for S6W. These distances are somewhat less than those calculated by Seabrook, t
6.2.2 Thyroid Dose Vs Distance The thyroid doses were not presented by Seabrook in the reviewed report but were discussed at some length in NUREG-0396. Hence, BNL calculated the risk of exceedi_ng the dose levels of 5, 25, and 300 rem to the thyroid, as was done in NUREG-0396 for Seabrook release categories S1W, S2W, and S6W. The results are given in Figures 6.2, 6.3, and 6.4. The results were truncated at 30 miles. The risk of exceeding 5 rem remained above 90% for all three release categories. The 300 rem curve shows a sharp drop at less than 10 miles for S2W and at about 15 miles for S6W. The curves were obtained by the same hazard function definition technique as discussed in Section 6.2.1.
6.2.3 Risk of Early Fatalities MACCS uses the hazard function approach to calculate early fatalities as discussed in NUREG/CR-4214 First, the cumulative hazard is calculated as:
H = In(2) (D/Dso) (6.1) where D is the dose and Dso is the dose required for producing an effect in 50% of the exposed individuals, and y determines the steepness of the dose effect curve. The fatality risk is then given as:
Risk = 1 - e-IN I+H2+H3 + Hg)
(6.2)
6-5 where H 3 is for red marrow, H 2 is for lungs, and H3 and Hg are for the lower large intestine- and small intestine. The risk is assigned 'a ' threshold of
'. 00 5.
In CRAC2 and CRACIT, the dose response is piece-wise linear due to irrad-1ation of the bone marrow, lung and GI tract. The' total risk is then:
R=R t + (1-R t)R 2 + (1-R t)(1-R 2)R3 (6.3) where R i , R2 , and R 3 are the risks to the three organs, respectively. MACCS gives somewhat. higher risk, principally because the lung dose is now consid-ered more effective in producing fatalities, and also because the hazard func-tion gives some risk at lower doses.
The effect of the code differences,is that MACCS predicts'a higher proba-bility of a small number of . deaths while CRAC2 predicts 'a higher probability for large numbers of deaths. This is shown in Figure 6.8 from a comparison calculation performed by Sandia National Laboratory for a severe ground level release. , .This is for a.. uniform population distribution without evacuation.
However, MACCS can predict substantially more early deaths when evacuation is modeled since the lung dose usually becomes dominant, but evacuation scenarios are not considered in this review.
BNL calculated the individual risk of fatalities versus- distances for S1W, S2W, and S6W as shown in Figures 6.2, 6.3, and 6.4. It can be seen that the risk is similar to the 200 rem curve, but is not directly correlated because of the nonlinear interactions in the above equations.
6.3 Comparisons of Results 6.3.1 Results of Seabrook Study j
BNL used as a basis for comparison the 200 rem risk of whole body dose as calculated by Seabrook using CRACIT and 200 rem red marrow risk using MACCS at BNL. This is the risk to a hypothetical individual located at a particular distance and actual population distribution is not considered. The BNL
?
. e 6-6 calculations were performed with meteorological data supplied by Seahrook.
Projected numbers of fatalities were not computed since BNL didn't have the actual Seabrook population data.
The Seabrook calculations for accidents S1W, S2W, and S6W are shown in Figures 6.5, 6.6, and 6.7 with the corresponding BNL results superimposed.
For S1W, MACCS predicts slightly higher risk at less than 8 miles and much lower risk beyond 12 miles. The differences may be partly explained by' i differences in weather sampling methods and plume rise formulations, since this is a high energy release. However, the differences in the tail of the i risk curve is not considered by BNL to be significant considering the overall uncertainties in the calculation.
For S2W, the MACCS code again predicts higher risk in close and a some-what sharper dropoff in the tail. However, the difference is that MACCS '
predicts a risk of .001 at 2 miles whereas the Seabrook results show this risk at 2.5 miles. !
For-S6W, the BNL results are close to those of Seabrook as in the case of S2W. MACCS predicts .001 risk at 4 miles whereas Seabrook predicts this risk of 200 rem at 6 miles.
In summary, BNL feels that the dose versus distance modeling is fairly presented by Seabrook and that the relatively small differences computed by RNL are probably explained by modeling techniques.
6.4 Sensitivity Studies 1
Two categories of sensitivity studies have been performed as part of the i
- BNL review. First, sensitivity calculations were performed to assess the affect of the duration of fission product release on the dose vs distance curves presented in PLG-0465. These sensitivity calculations are discussed in Section 6.4.1. Second, the impact on the dose vs distance and risk estimates in PLG-0465 of the various concerns raised by the BNL review was assessed in i
each section of this tectnical evaluation report. These revised risk esti-mates are summarized in Section 6.4.2.
i
- =
6-7 6.4.1 Sensitivity of Results to Multipuff Release BNL performed sensitivity calculations with regard to the multipuff releases and the release duration for S2W and S6W. The results are given in Figures 6.6 and 6.7, respectively using the one puff release categories defined by Seabrook. In both cases it was - found that a one puff release increased the risk and also that a shorter duration of the release (0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) further increased the risk. For the one puff, 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> release S2W, the 200 rem .001 risk distance increased from 2 miles to 7 miles and the S6W distance increased from 4 miles to 15 miles. This demonstrates that. a long release duration leads to a greater plume dilution and less. risk at larger distances.
Hence, in order to have confidence in the Seabrook calculations, one must have confidence that the releases will occur with rates and durations similar to those used by Seabrook.
6.4.2 Summary In the following sections, the BNL findings related to each review area are ,briefly summarized. BNL has attempted to follow the ground rules for comparison purposes discussed in Section 1 (namely mean frequencies for comparison with the safety goal and median frequencies for comparison with NUREG-0396 results), however, we have also attempted to include 1 discussion on the uncertainties associated with the risk estimates.
6.4.2.1 Interfacing systems LOCA A major concern resulting from the BNL review of the interfacing systems LOCA analysis in PLG-0465 and the RMEPS related to the determination of initiator frequencies. The effect of changing the initiator frequencies was determined by propagating the changes through the appropriate event trees in the RMEPS. The revised initiator frequencies resulted in the following changes to the frequencies of release categories S1W and S7W.
Mean Frequency Per Reactor Year Release Category PLG-0465 BNL Review S1W 4.0x10 8 1.4x10 7 S7W 6.3x10-8 1,ixig-6
~
6-8 l l
! 3 The above changes in release category frequencies have no impact on indi-vidual risk of early fatalities if no evacuation or 1 mile . evacuation is assumed. This is because release category S2W dominates this risk neasure, and it has a frequency of 2x10 5 Only when ~ a 2 mile evacuation is assumed (and the early fatality risk for category S2W becomes zero) do the above :
changes in release category frequencies change the original PLG-0465 esti-
- mates. However, with a 2 mile evacuation the,early fatality risk is very low j and well below the safety goal. The 200-ren dose versus distance curve in
! PLG-0465 10 is also not influenced by the above changes in release category frequency. This is because only release category S1W has a significant proba-bility of exceeding a 200-rem dose, and the revised probability of this cate-
! gory is not sufficiently high for it to appear in 'the PLG-0465 dose versus f distance curves.
1 There is of course uncertainty associated with predicting the frequency i
) of interfacing systems LOCAs. However, the frequency of interfacing systems
} LOCAs resulting in release category S1W would have to increase by two orders i of magnitude before the Seabrook dose versus. distance curves would approach l the curves given in NUREG-0396. One can therefore conclude that interfacing
! systems LOCA are unlikely to increase the risk profiles presented in PLG-0465 I to the level presented in NUREG-0396. This is not too surprising because when
! no evacuation is assumed, the higher frequency events dominate risk and inter-i facing systems LOCAs did not contribute to the dose versus distance curves
! taken from NUREG-0396.
i 6.4.2,2 Accidents during shutdown This topic was not originally addressed in PLG-0465 and a detailed i assessment of such events is beyond the scope of the current BNL work on this
! project. However, the applicant was requested to provide information on the
, risk associated with accidents during shutdown. The results of the appli-t l cant's assessment of such accidents were presented in the form of sensitivity studies in a draft version of this report. The applicant provided additional I frequencies to the existing release category frequencies given in PLG-0465 to f assess the impact on risk from accidents during shutdown. A base case and a I bounding case were presented by the applicant. The additional frequencies i
. ._ _ _ _ _ . . _ . . ~. __ ._ - - _ _ _ _ .
6-9' associated with these accidents are given below:
k Mean Frequency Per Reactor Year Release Category Power Base Case- Bounding j Operation Events Shutdown Events Shutdown Events S.5 1.1x10 1.7x10 5 S.2- 2.1x10 4.9x10 7 S.6 6.5x10 7 .7.1x10 8 5x10 6 BNL was _ not in a _ position to assess the above frequencies for these
!- events because there remained fundamental questions regarding ,the modeling of' these scenarios. However, in spite of this, the applicant's results' were included in the draft report for comparison with the B'NL sensitivity. study results on other topics. It should be noted that the applicant considered the upper bound estimates to be very conservative. In particular, in order to
- assess the impact of these events, they were included in source, term cate-gories derived for accidents from full power, which could lead to predicts of
- shorter times and larger quantities of fission product aelease than would be expected from accidents during shutdown.
In a subsequent submittal by the applicant, the consequences of accidents from shutdown were revised. The applicant felt that 94 percent of accidents j
at shutdown would occur at times later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after scram. Thus, the consequence estimates were reanalyzed assuming release times of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
The later release times resulted in dose versus distance curves which fall off l at much shorter distances from the site boundary than the original dose versus
[ distance curves. BNL has checked this result and confirmed that if the release does occur at times greater than 48 hars, then the new dose versus distance curves are reasonable.
4 I
The results of the latest applicant's assessment of accidents at shutdown are reproduced in Figure 6.9. As noted above, a detailed assessment of such events ~is beyond the scope of the current BNL review. However, based on our limited review of the applicant's assessment of these events, we still have reservations about the results. These reservations are discussed in greater I - . _ - . - . _ . - - - - - - - _ - _ . - _ _ _ _ -
6 detail ~ in the body of this report but until they are resolved, we are unable to assess the validity of tha risk estimates presented by the applicant in Figure 6.9 4
'6.4.2.3 Induced steam generator tube rupture For accidents in which the primary system is at high pressure during core
- uncovery and melting, it is possible that 'large natural circulation flow patterns could develop within the primary system. . These flow patterns could in turn heat-up and degrade regions of the primary system remote from the -
reactor core. Of particular concern is the possibility of-degrading the steam generator tubes such that the primary system could become open to the secon-dary system. If -the secondary system were in turn open to atmosphere, then a direct path could exist between the primary system and .the atmosphere, which bypasses containment. This topic was not included as part of the work scope
' for the current BNL review. However, the topic has been. reviewed in detail by the NRC staff and is the subject of continuing NRC and industry research j activities. Therefore, BNL performed simple sensitivity studies to assess the potential impact of induced steam generator tube rupture on risk at Seabrook.
The results of -the sensitivity study are given in Figure 6.10.
Because of uncertainty in predicting events of this nature and the fact that BNL did not evaluate this issue in detail, we were not able to develop a j best-estimate frequency for induced SGTR. The sensitivity study therefore presents a range of possible frequencies for induced SGTR. The frequency of high pressure squences and the conditional probabilitiec of failure of the I
operators to depressurize and induced SGTR that were used in the sensitivity study are given below:
I 4.0x10 5 x 0.5 x 0.3 = 6.0x10 6 per reactor year I
4.0x10 5 x 0.2 x 0.01 = 8.0x10 8 per reactor year.
In order to estimate the impact of the above probabilities on risk, an appropriate source term category had to be selected. It was decided to allo-cate SGTR events to release category S1W, which represents a large early bypass of the containment. It was felt that this was a conservative
. , . - - - , . = , - - - - . _ _ - -,m_.-_,-,_,m.,,--.,w.,.,__-., . . . - _ - _ - --.-.-,,_-,w_--_. _,_,----,,,,-_,e ---,m,,_ . - - . _ _ , , - - , - - . . . - , -
6-11 I
assumption because significant retention of the fission products in the secondary side- could occur and this was not considered when calculating the S1W release fractions. The impact of adding the above frequencies to source term category S1W is illustrated in Figure 6.10.
The lower estimate of the frequency of induced SGTR has no impact on the risk estimates presented in PLG-0465. The higher estimate of the frequency of induced SGTR has no influence on the individual risk of early fatalities within 1 mile . of the site boundary if no evacuation is assumed but does influence the 200-rem -dose versus distance curves as shown in Figure 6.10.
Allocating the probabilities of- SGTR events to rel_ ease cat'egory S1W has the largest. impact on the dose versus distance curves. However,'the impact on the risk of early fatalities within 1 mile is negligible because S1W has very little risk of fatalities within this distance (refer to Section 1). If the probabilities of SGTR events were added to release category S6W, the impact on the dose versus distance curves would be less but the risk of fatalities within 1 mile would increase slightly if no evacuation is assumed.
It should be noted that the range of frequencies used for the induced SGTR sensitivity study were developed to cover our lack of understanding in this area and that the NRC staff believes that the actual probability of a SGTR is closer to the lower estimate. However, one reviewer 11 of the BNL draft report felt SGTR to be a potentially more "significant" issue than was implied in our evaluation. It was not BNL's intention in the draft report to minimize the potential importance of this issue, and the range we presented did not represent an upper bound. It was an attempt to reflect the best judgments of several experts on a very difficult subject. There is a great deal of uncertainty associated with predicting such events and it is therefore
- prudent to indicate the impact on risk of a range of assumptions, i
6.4.2.4 Containment isolation failure and pre-existing leakage l The applicant's assessment of pre-existing 14akage and containment isola-tion failure was reviewed by the NRC staff. Based on the NRC staff review of the information available, it was concluded that the purge and vent valves in a fully closed configuration should provide reliable isolation of the Seabrook r
I
6-12 containment under severe accident conditions up to the pressure corresponding to 1 percent. hoop strain'in the containment. _
The NRC staff also concluded that the applicant has presented a reason-able approach for the consideration of pre-existing leaks, both small - and 1arge. The approach adopted ' by the. applicant wast to use information on containment unavailability developed in ~ a study 12 by the Pacific Northwest Laboratory (PNL) to assess the impact on risk of pre-existing leakage. The -
applicant used ' this information to bound the effects ' of the data in the PNL
. study (NUREG/CR-4220) even though they considered that it did not apply to
. Seabrook. The results of the applicant's assessment are given in Figure 6.11. From an inspection of Figure 6.11, it is apparent that the impact of i the NUREG/CR-4220 data on the dose versus distance curves in PLG-0465 is not great..
6.4.2.5 Containment structural capacity
,- Based on its nonlinear finite element analysis of the Seabrook contain-
! ment, BNL concluded that a shear failure at the base of the cylindrical wall l 1s a potential failure mode but would not occur before reaching a pressure of 165 psig.
BNL agrees that the containment structure would reach a general yield
! state in the hoop reinforcing steel at a pressure of 157 psig and that it is 4
appropriate to consider this pressure as a lower bound pressure for the hoop l mode of-failure. However, BNL believes that the median hoop failure pressure should correspond to the one percent strain level in the hoop reinforcing steel, which is a pressure of 175 psig. The above pressures are for the wet l containment conditions. For the dry containment conditions the corresponding
{ median failure pressure is 158 psig and the lower bound pressure (general yield) is estimated to be 145 psig. This latter value is based on the reduc-l tion factor recommendation in Section 11.3.4.1 of PLG-0300 With regard to containment penetrations, RNL believes that the failure pressures should be based on containment deformations assuming no bond strength. between the reinforcing steel and concrete. Based on this assumption i
6-13 BNL estimates median failure pressures for the wet containment condition of 159 psig and 167 psig for two critical penetrations. For the penetration with the lower failure pressure, BNL agrees that a Type A (less than 6 square inches) leak path is appropriate for the median estimate; however a Type B (6 square inches to about 0.5 square foot) leak path should be considered as an upper bound estimate. For the penetration with the higher failure pressure, BNL agrees that a Type B leak path is appropriate for the medium estimate; however, a Type C (greater than 0.5 square foot) should be considered as an upper bound estimate.
For the dry containment conditions, BNL estimated the median failure pressures for the above two critical penetrations to be 147 psig and 152 psig, respectively. These values are also based on the reduction factor re' commended in Section 11.3.4.1 of PLG-0300.
Although BNL has performed some independent calculations to support its conclusions regarding the containment strength, it also relied on the results of calculations performed by PSNH and its contractors. Therefore, BNL recom-mends that a complete and independent check of all relevant containment strength calculations be performed by PSNH. PSNH committed to such a check in their letter to the NRC dated October 31, 1986 and has indicated that such a check has been completed.
6.4.2.6 Containment loads BNL's assessment of the capacity of the Seabrook containment (described above) has to be combined with severe accident loads (pressure / temperature histories) to determine the potential for early containment failure. BNL does not have Seabrook-specific containment loads and was not able to generate such loads given the limited scope of the current review. However, BNL has been involved in updating (NUREG/CR-4551,13 Volume 5) the risk profile for the Zion plant for input to the NRC's " Reactor Risk Reference Document," NURE'i-1150.
The updating of risk for Zion was based on a methodology developed as part of the Severe Accident Risk Reduction Program (NUREG/CR-4551, Volumes 1-4) at Sandia National Laboratory (SNL). This methodology used expert judgnent in an attempt to estimate the uncertainty associated with determining containment
6-14 -
loads. The methodology was developed at SNL specifically for the Surry plant but was extrapolated to Zion at BNL The Zion plant is _very similar to Sea-brook in terms of the containment volume to reactor power ratio. Thus, extrapolation of the Zion loads to Seabrook would give some indication of the impact of applying this new methodology to Seabrook. It must be emphasized that this exercise should in no way be interpreted as a Seabrook-specific cal-culation. It simply gives some indication of the sensitivity of the Seabrook results to the types of uncertainty in estimating containment loads discussed in NUREG-1150. It should also be noted that this work is preliminary and has not yet undergone full peer review outside of NRC and its contractors. It is, therefore, subject to revision.
The range of containment loads reported in Volume 5 of NUREG/CR-4551 for Zion is very wide and far exceeds the loads that would be considered credible by the applicant for Seabrook. Of particular interest is the loads at the time of reactor pressure vessel failure. These loads can range from about 60 psia to 200 psia depending on whether core melt is occurring with the primary system at high or low pressure and on whether or not containment heat removal systems, CHRS (sprays and fan coolers) are operating. The higher containment loads are postulated to occur for accidents in which the primary system pres-sure remains high immediately before reactor pressure vessel failure. For these accidents, direct heating of the containment atmosphere by core debris or hydrogen combustion with a steam spike at the time of reactor vessel fail-ure are possible mechanisms for failing the containment. The applicant has presented information which indicates that these mechanisms are not credible ways of failing the Seabrook containment. However, as noted above, BNL does not have Seabrook-specific containment loads so we,cannot, at this time, elim-inate these mechanisms as potential ways of failing the Seabrook containment.
For accidents with the primary system at high pressure and without the CHRS operating an approximate median load of 135 psia (120 psig) was predicted for Zion. If this median load is compared against the capacity of the Sea-brook containment given by the BNL review, one would conclude that the poten-tial for early containment failure at Seabrook is very low and would not influence the risk estimate in PLG-0465. However, the range of loads esti-mated for Zion implies considerable uncertainty. The 95th percentile estimate
6-15 i of the probability of early containment failure at Zion is quoted as 0.17 in Volume 5 of NUREG/CR-4551. If this early containment failure probability were also true for Seabrook, the risk estimates in Figures PLG-0465 would increase ;
significantly. However, the capacity of the Seabrook containment is greater than Zion (the general yield for Seabrook is 157 psig compared with 134 psig for Zion) so the 95th percentile estimate of early containment failure should be lower at Seahrook than Zion. However, BNL cannot at this time quantify how much lower because we have not quantified Seabrook-specific containment event trees with Seabrook-specific containment loads combined with our estimate of the structural capability of the Seabrook containment.
6.4.2.7 Source terms The fission product source terms used in PLG-0465 10 were reviewed in terms of their consistency with the approaches used in WASH-14002 and found to be appropriate. A misprint in PLG-0465 related to the release of noble gases for release category S2W was discovered. However, correcting'the noble gases release was found to have no impact on the risk profiles in PLG-0465. In addition, the argument presented by the applicant that water in the residual heat removal (RHR) vault is sufficiently subcooled to warrant consideration of significant decontamination was found to be reasonable. This is an important consideration for the subset of interfacing systems LOCAs where the break location in the RHR line is low in the RHR vault. Under these circumstances, with the break location submerged considerable scrubbing of the aerosol fission products would occur.
This would result in much lower aerosol fission product release than for accidents in which the break location was uncovered.
The applicant considers the WASH-1400 source terms used in PLG-0465 to be very conservative and the applicant has high confidence that the source terms would not be exceeded in a real accident. BNL found the source terms used in PLG-0465 to be consistent with WASH-1400 methodology but we are not as con-fident as the applicant that they could not be exceeded. The new source term methods (refer to NUREG/CR-4551, Volumes 1-5) indicate that if the containment fails late or if there is graduai leakage from containment then the aerosol fission product release is likely to be lower than would be predicted by WASH-1400 methods. This is because WASH-1400 methods underpredicted aerosol
6-16 agglomeration and settling. Therefore, if the new methods were applied to release categories S2W and S6W, the predicted aerosol release would be lower than WASH-1400 values. However, the new methods also indicate that if con-tainment fails early and the opening is large, then i;here is considerable uncertainty associated with predicting fission product release. The uncer-tainty ranges associated with fission product release in NUREG/CR-4551 can, for certain accident sequences and early containment failure modes, exceed the WASH-1400 predictions. This uncertainty would principally affect the S1W release category at Seabrook.
6.4.2.8 Consequence Model The applicant used the CRACIT code for their consequence assessments in PLG-0465. BNL compared CRACIT predictions of dose versus distance with predictions from the MACCS code, which was developed at Sandia National Laboratory (SNL) under NRC sponsorship. The comparison of the dose versus distance curves for the CRACIT and MACCS codes was reasonably good. There-fore, BNL feels that the dose versus distance modeling in PLG-0465 is fairly presented and that the relatively small differences between CRACIT predictions and those computed by BNL using MACCS are explained by differences in modeling techniques used in the two codes.
BNL could not check the risk of early fatalities reported in PLG-0465 because we did not have the population distribution around the Seabrook site.
Therefore, as BNL had only CRACIT results for early fatalities, it was decided to use CRACIT results for both early fatality risk and dose versus distances in the BNL sensitivity study. This was done simply so that we had consistency between the two risk measures and not (as implied by the applicant's review of the draft BNL report) to present more " conservative" CRACIT results. We found that CRACIT in general predicted dose versus distance curves that extended further than the MACCS code and in this sense CRACIT is more " conservative" than MACCS. However, we note that MACCS predicts more early health risk than CRACIT and therefore the use of the CRACIT results is probably not "conserva-tive" for this risk measure. In addition, CRACIT predicted that only 9 early fatalities would occur within 1 mile of Seabrook if release category S1W was to occur compared with an estimated total early fatality risk of 746 (refer to
~r o
' 6-17~
L Table 1.2). The reason for the very small risk 'of early fatalities'close to the Seabrook site boundary for the S1W release category is probably due to the plume lift-off model in CRACIT.
The S1W release category has 'a high energy plume and CRACIT would calculate significant elevation of the S1W plume rela-tive to the other release categories. ~
This reduces the probability of early fatalities close to the site but increases the probability of early fatalities at larger distances. . Also, the weather sampling in1 CRACIT - can result in increased risk of early fatalities at greater distance from the site depending on when rain is predicted to occur. Sensitivity studies using the MACCS code at BNL indicate much less sensitivity to high energy releases than CRACIT. It is therefore likely that if BNL had calculated early fatalities for the S1W release category using MACCS with the actual Seabrook population distribution we would have calculated mo're Learly fatalities closer to the Seabrook site boundary than predicted by CRACIT. This statement is supported by the com-parison of the CRACIT and MACCS 200 rem dose versus distance curves in Figure 6.5, which indicate that MACCS has a higher probability of exceeding this dose level cose to the site boundary than CRACIT.
This difference in MACCS and CRACIT predictions has important implications for. comparison. with the safety goal, which deals with the risk of early fatalities in the population within ene mile of the site boundary. A major conclusion of the BNL sensitivity studies is that the safety goal compariosn is relatively insensitive to uncer-tainties in estimating the frequency of release category S1W. However, this conclusion is based on the applicant's CRACIT calculations, which predict a very low risk of early fatalities close to the site boundary. It is not clear that if BNL had performed MACCS calculations and used.these' predictions in the
~
BNL sensitivity studies that the safety goal comparison would have been so favorable.
This is an area that requires further investigation in any follow-on effort.
6.5 References .
- 1. Collins, H. E.,
et al., " planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants," prepared for the U.S. Nuclear Regulatory Commission, NUREG-0396, December 1978.
6-18 a-
- 2. U.S. Nuclear Regulatory Commission,." Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," WASH-1400, NUREG-75/014, October 1975.
- 3. Reactor Safety Study ' Consequence Model, " Computer Code Users Manual,"
U.S. Nuclear Regulatory Commission, undated.
- 4. Ritchie, L. T., et al ., " Calculations of Reactor Accident . Consequences Version 2.CRAC2: Computer Code Users Guide," NUREG/CR-2326, February 1983.
- 5. Pickard, Lowe, and Garrick, Inc., "Seabrook Station Probabilistic Safety Assessment," prepared for Public Service Company of New Hampshire and Yankee Atonic Electric Company, PLG-0300, December 1983.
- 6. Chanin, D. I., Ritchie, L. I., and Alpert, D. J., "MELCOR Accident Conse-quence Code System MACCS User's Guide," Sandia National Laboratories ,
Albuquerque, NM, to be published.
- 7. Evans, J. S., et al., " Health Effects Model for Nuclear Power Plant Consequence Analysis," NUREG/CR-4214, July 1985.
- 8. BEIR-1 Report (1972): The Effpcts on Populations of Exposure to Low Levels of Ionizing Radiation. Report of the Advisory Committee on the Biological Effects of Ionizing Radiation. Division of Medical Sciences, National Academy of Sciences, National Research Council, Washington, DC.
- 9. BEIR Report (1980): The Effects on Populations of Exposure to Low Levels of Ionizing Radiation. Report of the Advisory Committee on the Biologi-cal Effects of Ionizing Radiation. Division of Medical Sciences, Nation-al Academy of Sciences, NationL1 Research Council, Washington, DC.
- 10. Pickard, Lowe, and Garrick, Inc., "Seabrook Station Emergency Planning Sensitivity Study," prepared for Public Service Company of New Hampshire and Yankee Atomic Electric Company, PLG-0465, April 1986.
- 11. Theofanous, T. G., " Review Comments," University of California, Santa Barbara, dated January 12, 1987.
- 12. Pelto, P. J., et al., " Reliability Analysis of Containment Isolation Systems," NUREG/CR-4220, PNL-5432, June 1985.
- 13. " Evaluation of Severe Accident Risks and Potential for Risk Reduction,"
NUREG/CR-4551, Volume 5, February 1986.
l
6-19
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6-20
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6-23 te' 1$
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Figure 6.5 Dose versus distance curve for release category S1W from Seabrook for no immediate protective action with BNL results using MACCS superimposed,
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3-Puff 200 rem
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=
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i
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- 5:18 E
,e < >
is 1 ~ 't es 'is t tea DISTANCE (it!LES)
Figure 6.6 Dose versus distance curve for release category S2W from Seabrook for no immediate protective action with BNL results using MACCS superimposed.
S r
t 2
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3 Puff 0.5 hr E
l[jj RE"
- ,'jj+}56.06 k duration
-4 16 ' -( -' ' \-
16
-4 10-1 '168 ~ '161 162 O! STANCE (MILES)
Figure 6.7 Dose versus distance curve for release category 56W
, from Seabrook for no immediate protective action with BNL results using tiACCS superimposed.
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Figure 6.8 Comparison of MACCS to CRAC2 cods -(fNm e preliminary benchmark calculations performed '
by Sandia Labs). '
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<h ; release, PSNH letter (SBN-1225) -
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om ' l1 \ - Shutdown events assuming 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> li release, PSNH letter (NYN-87-002) -
"O 1
\ -
i l\ g 200 REM o.001 ' ' ' ' ' ' '' ' ' ' ' ' ' ' '
1 10 100 1,000
, . OlSTANCE (MILES) ,
Figure 6.9 Comparison of 200 rem-dose versus distance curves for conservative assumption of no credit for operator recovery of open equi culations performed by PSNH.) pment hatch.(Cal-
,p s: y ...
4 6-28
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- I b regarding SGTR (optimistic assumps _
,. l 3
tion off' graph) 0.001 ' ' ' I I ' ' ' i' I b '
t, 1
..J 10 '
100 ' ' ' ' ' 1,'000 DISTANCE (MILES)
,. Figure 6.10 Comparison of BNL sensitivity studies with PLG-
)
0465 and NUREG-03'J6. (200-rem immediate protective actions.) plots with no N
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6-29 1
i i i iiii61 e i i i i i i n g, . . . . ..i,
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[g 0.1 -
ou NUREG-0396
- I g
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aso; -
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aw
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sh ,' l Effects of conservative _
9 -
interpretation by PSNH of
$w9 -
h NURtG/CR-4220 data,.PSNH gy -
l\ letter. _
uo g t ~
t l
l 0.001 I ' 'I i I t i t i iil i i i iiii e
! 1 10 1og 1,000 l DISTANCE (MILES)
Figure 6.11 Comparison of 200 rem-dose versus distance curves
! for conservative interpretation by PNSH of NUREG/CR-4220 data. (Calculations performed by PSNH.)
l i
t I
Table 6.1 Summary of Release Categories Representing Hypothetical Accidents (from the RSS)
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