ML20112H346
| ML20112H346 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 03/26/1985 |
| From: | Hodder M CENTER FOR NUCLEAR RESPONSIBILITY, HODDER, M.H. |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#285-341 84-496-03-LA, 84-496-3-LA, OLA, NUDOCS 8504020227 | |
| Download: ML20112H346 (4) | |
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,f5If' UNITED STATES OF AMERICA .-
NUCLEAR REGULATORY COMMISSION
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-250-OLA
) 50-250-OLA Florida Power J Light Company )
) ASLBP No. 84-496-03 LA Turkey Point Plant )
Units 3 & 4 )
)
INTERVENORS' MOTION TO SUSPEND OR REVOKE LICENSE AMENDMENTS Intervenors motion that the Atomic Safety and Licensing Board
( Board) suspend or revoke license amendments No. 93 and No. 99, issued for the Turkey Point nuclear reactors on December 23, 1985, until the Board has determined that a valid technical basis, including but not limted to BART computer code REFLOOD model, has been established.
On March 18, 1985, Counsel for the Florida Power & Light Company
( FPL ) sent a letter to the Atomic Safety and Licensing Bcard which informs that there is no valid technical basis for the WREFLOOD BART computer model. (See letter from Micheal A. Bauser, FPL, to the Atomic Safety and Licensing Board, March 18, 1985, Attatchment A).
From the outset of these proceedings, Intervenors have contended that the Nuclear Regulatory Commission ( NRC ) and the utility appear to be experimenting with an untested and unproven technology which threatens catastrophic loss in terms of health and safety of the residents of South Florida.
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The Intervenors 'in their original pleadings and in the affidavit of Dr. Gordon Edwards pgs. 6-7 (see attatchment B) disputed the adequacy of the BART computer model used as a basis for the license amendments.
The FPL letter,taken with Intervenors' contentions, not only clearly identifies at least one disputed issue of material fact, but undermines the Commission's own determination that there was a no_significant safety hazards consideration involved, which was the basis for issuance of the license amendments and the decision that no prior public hearing was required. There is a strong likelihood that the deficiencies in the BART computer model could have been established prior to issuance of the license amendments if the Commission had required a prior public hearing.
Without a valid technical basis for the subject license amendments, Intervenors submit that the . Board must now suspend or revoke these license amendments, with the requirement that the Licensee, Florida Power & Light Company, operate the facility in accordance with_the original technical specifications and safety margins established in the original operating license, and until this Board has determined that there exists a computer model that allows operation of the plant widdn the requirements of 10 C.F.R. 50.46 and 10 C.F.R. Part 50 Appendix K.
Intervenors suggest that the Board establish an expedited briefing schedule so that the parties may address, and the Board resolve, this important and and unresolved safety issue promptly;
'(3) since presently there is no valid legal, technical, or mathematical basis for operation of the Turkey Point nuclear power plants under-.the subject license amendments.
Respectfully submitted, l /' /f '
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-Q Martin H. Hodder Attorney for the Center for Nuclear Responsibility and Joette Lorion 1131 NE 86 Street Miami, F1. 33138 (305) 751-8706 Dated: March 26, 1985
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c-l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-250-OLA
) 50-251-OLA Florida Power & Light Company )
) ASLBP No. 84-496-03 LA Turkey Point Plant )
Units 3 & 4 )
)
CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing "Intervenors' Motion to Suspend or Revoke License Amendments" have been served on the following parties by personal service at the ASLB Prehearing Conference on March 26, 1985, in Miami.
Dr. Ibbert M. Iazo, Chairman Docketing & Service Section
- Atomic Safety & Licensing Board Panel U.S. Nuclear Regulatory Qxtmission U.S. Nuclear Regulatory Comnission Washington, DC 20555 >
Washington, DC 20555 Dr. Btmeth A. Inebke Harold F. Reis, Esquire Atomic Safety & Licensing Board Panel Naman & Holtzinger PC U.S. Nuclear Regulatory Ommission 1615 L. Street NW Washington, DC 20555 Washington, DC 20036 Dr. Richard F. Cole Norman A. Coll, Eqsuire Atomic Safety & Licensing Board Panel Steel, Hector & Davis U.S. Nuclear Pegulatory Omnission 4000 SE Financial Center Washington, DC 20555 Miami, F1. 33131-2398' Mitsy Young, Esquire d (X.A)$ 6 0-Office of General Counsel U.S. Nuclear Regulatory Cbmnission Martin H. Hodder *
- Washington, DC 20555 Attorney for the Center for Nuclear Responsibility and Joette Iorion 1131 NE 86 Street Miami, Fl. 33138 (305) 751-8706 '
- signed in his absence DATED: March 26, 1985 by Joette Lorion
- Docketing & Service was served by mail.
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Dr. Robert M. Lazo, Chairman Dr. Richard F. Cole Administrative Judge Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board
- U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington,-D.C. 20555 Dr. Emmeth A. Luebke Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission l.
Washington, D.C. 20555
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via Messenger 4
Re: In the Matter of ~
Florida Power and Light Company (Turkey Point Plant, Unit Nos. 3 and 4)
Docket Nos. 50-250 OLA-1, 50-251 OLA-1
Dear Members of the Board:
Last Friday, March 15, 1985, Florida Power & Light
! Company (FPL) was informed by Westinghouse Electric Corporation that, on March 9, l'985, it was determined necessary to revise the
- ECCS evaluation model procedure by which the core flooding l rate information generated by the WREFLOOD code is introduce'd into the BART heat transfer coefficient calculation. This change was made after Westinghouse discovered, while performing work unrelated to Turkey Point Units 3 and 4, that the procedure for introducing the flooding rate information into the BART code allowed information inconsistent with the WREFLOOD results to be used, thereby resulting in lower calculated peak clad temperature (PCT) than otherwise would have been calculated in some cases. FPL has been informed that preliminary ECCS performance reanalysis utilizing the revised procedure results in a calculated PCT higher than the 1972 *F calculated PCT indicated in the August 10, 1984 Siu 3 AUu63E
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" Licensee's Motion for Summary Disposition of Intervenors' Contention (b)," and the " Affidavit of Mark J. Parvin" attached by 10 CFR thereto, S 50.46. but well below the 2200 'F limit established Reevaluation and parties informed. is continuing and we will keep the Board Sincerely, M
. Michael A. Bauser MAB:cw l
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CANADA PROVINCE OF QUEEEC DISTRICT OF M0tTTREAL.
t AFFIDAVIT OF GORDON D. J. EDWARDS '
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' I, Gordon Edwards, being duly sworn, say as follows:
- 1. I am President of the Canadian Coalition for Nuclear Respon-sibility (Inc.) and Professor of Mathematics and Science at Vanier College (Montreal). My business addresses are, respectively, P.O. Box 236, Snowdon Post Office, Montreal, Quebec, Canada, H3X 3T4; and 5160 l
Decarie Boulevard, Montreal, Quebec, Canada, H3X 2H9. A summary of my professional qualifications and experience is attached hereto as Exhibit A, which is incorporated herein by reference. I have personal l knowledge of the matters stated herein, and believe them to be true and correct. This Affidavit is offered in support of Ms. Joette Lorion's Intervention, specifically contentions (b) and (d), in the l
matter of Florida Power and Light (the Licensee), concerning the Turkey Point Nuclear Generating Station, Units 3 8e 4.
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- 2. To support the Licenaco'o integrated progrcm for vaccal flux reduction, in order to resolve the pressurized thermal shock issue,
( the Licensee has decided to change the core configuration -- moving from the present regime involving LDPAR (" low parasitic") fuel to a new regime involving DFA (" optimized fuel assembly") fuel. During the transition period, a " mixed core" of LDPAR fuel and DFA fuel will be used. At the same time, Licensee has applied for permission to offset the reduction in volume of fuel by increasing the hot channel Fgg limit from 1.55 to 1.62, and by increasing the total peaking factor F, limit from 2.30 to 2.32, thereby obtaining increased power from the lower fuel density. This change in design will increase the fuel temperature in certain parts of the reactor core.
- 3. Running the fuel at a hotter temperature materially increases both the probability and the consequences of cladding failure, whether there is a loss-of-coolant accident (LOCA) or not. This increase in probability results from the fact that the margin of safety is mate-rially reduced by operating the fuel at a higher temperature. At a certain critical temperature, if there is no LOCA, an insulating film of steam will begin to form around the hot cladding surrounding the fuel (" departure from nucleate boiling" or "DNB"), thereby sharply reducing the rate of heat transfer to the coolant, leading to over-I heating and rupture of the cladding, which in turn releases fission products (mainly radio-iodines and radioactive noble gases). If there is a LOCA, the same kind of damage can be done by failing to rewet the fuel fast enough. In either case, the time required to reach the critical temperature (f ollowing an abnocmal_ occurrence) will be shorter if the initial temperature of the fuel is higher. Moreover, the amount of radio-iodine available to be released in the event of a l
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cladding failure is directly related to the operating temperature of the fuel prior to the cladding " failure; by running the fuel at a
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higher temperature, more radio-iodine is in the " gap" between the cladding and the fuel, and theref ore the consequence of a cladding failure is worse (in turns of the quantity of fission products available to be released).
- 4. The Nuclear Regulatory Commission has acknowledged that there will be a " reduction in safety margin resulting from the increase in the F 6H and Fg limits" [ Federal Register, Vol.48, No. 196, page 45862], but argues that there are "no significant hazards" involved in
! granting the Licensee's request. The basis f or this ~ judgment is a safety evaluation carried out by the Licensee which calculates peak clad temperatures following hypothetical LOCAs to be "within the maximum limit of 22OO*F", and which also calculates that the
( " departure from nucleate boilin:) ratio" or "DBNR" has an additional f
margin of safety that had previausly not been identified. *
- 5. Licensee's Motion for Eammary Disposition of Intervenor's Contention (b) states, in part:
Section 50.46 of Nuclear Regulatory Commission regulations requires that an ECCS analysis be performed with an acceptable evaluation model, and result in a calculated maximum fuel element cladding temperature not greater than 22OO'F....
ECCS evaluation model analysis utilizing the BART code results in a calculated fuel rod peak clad temperature (" PCT") of 1972*F for a homogenous core of either low parasitic ("LOPAR") fuel or optimized fuel assembly ("DFA") fuel. However, in the current period of transition, when mixed cores of LDPAR and DFA fuel are utilized at Turkey Point, the analysis results are slightly effected Csic3 by the fact that the hydraulic resistance of DFA fuel is 4.5% higher than for LDPAR fuel. This
... results in approximately 10*F increase in PCT over the calculated 1972*F PCT for a homogenous core, which is well within the 22OO'F criterion....
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- 6. Licensee's Statement of Material Facts as to Which There is No Genuine Issue to be Heard adds the followings
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ECCS analysis has also been performed for a homogenous core with the previously approved evaluation model utilizing the Westinghouse Full Length Emergency Cooling Heat Transfer (FLECHT) correlation, resulting in an indicated peak clad temperature of 2130*F. A 10*F increase in temp-erature due to a mixed LDPAR and DFA core also results in a PCT less than the 22OO'F limit....
- 7. Because of the extreme conditions prevailing in the event of a loss-of-coolant accident (LDCA), there are no mathematical models in existence which can accurately model the complex interactions between fuel, cladding, steam, and water which will take place during reflood.
Classical mathematical analysis depends on assumptions of continuity which are clearly violated under accident conditions. There is no accepted body of mathematical knowledge which allows one to cope reliably with discontinuities such as those encountered in a LDCA-reflood sequence. Under such conditions, the mathematical models and computer codes developed to analyze conditions in the reactor core are, at best, vastly oversimplified conceptualizations based on the
" average" behaviour of an " ideal" system (existing only in the mind) enjoying such desirable properties as uniformity, symmetry, and pre-dictability -- properties which do not necessarily correspond to reality. In short, accident analysis is an art with scientific under-pinnings, but it is not an exact science. It predicts, not the precise behaviour of the system under study, but the " expected" or
" probable" behaviour of the system under study. There is always a certain tnquantifiable chance that the analysis itself is funda-mentally wrong: wrong not just in its conclusions, but in its underlying assumptions.
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- 8. The history of nuclear power is rife with examples of
,( occurrences having major safety implications which were not predicted '
by prior mathematical analysis; for examples (a) The 1966 accident at the Fermi nuclear reactor near Detroit was considerably worse than the " maximum credible accident" foreseen during the licensing process. Indeed, the accident involved the blockage of three fuel channels (with consequent overheating of the fuel) by a triangular piece of sheet metal whir.h was nowhere indicated in any of the engineering specifications of the plant; the metal triangle had been added in the core area as an af ter-I thought by a construction foreman.
The mathematical probability of the Fermi accident, prior to the onset of the accident, was "zero". No prior mathematical ana-lysis could have taken into account the existence of the metal triangle, of which there was no record. In retrospect, of course, the mathematical probability of the Fermi accident can now be seen as quite high; indeed, very close~to "one". Since the metal triangle was improperly welded, it was bound to come loose sooner or later; once loose, it was bound to obstruct the flow of coolant through the coreg hence the accident.
An important lesson can be drawn from this examples mathematical probability is, to a large extent, a measure of our ignorance.
The calculation of probability is aff ected by new inf ormation.
i (b) The Three Mile Island accident offers many similar examples of events of " probability zero" which have actually occurred --
I such as the simultaneous unavailability of all f eedwater pumps, l the onset of emergency cooling without a pipe break, and the initiation of an extremely serious accident because of the malfunction of a non-safety-related valve.
(c) Here in Canada, the sudden rupture of a pressure tube in the core of the Pickering Nuclear Generating Station (Unit 2) last summer was not only unforeseen, but was in direct contradiction to earlier mathematical analysis which had predicted a " leak-bef ore-break" phenomenon. The tube suffered a longitudinal crack about one meter in length, intersected at right angles by a 270-degree circumferential crack, without any prior leakage.
(d) In the case of the Turkey Point Nuclear Generating Station, the degradation of the plant's steam generators to the point where replacement became necessary would not have been predicted by prior mathematical analysis, nor was the problem of pressure vessel embrittlement foreseen when the plant was first licensed.
In fact, nuclear technology is perhaps unique in becoming more and more complicated as time goes on, rather than more and more simple.
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The number of nnw, fundam;ntelly important, previoucly ovcrlooksd, still unresolved problems is greater in the nuclear field than in I
( almost any other field of technology which has been commercially available for a comparable period of time.
- 9. Under the circumstances, it would be unwise to relax already existing safety margins solely on the basis of mathematical analysis using computer codes which predict that cladding failure will not occur in the event of LOCA, or that DNB will not occur otherwise. The following points should be borne in minds (a) The Licensee has just faced one major unanticipated problem by replacing the steam generators, and is now facing another major unanticipated problems potential pressure vessel embrittlement.
Until these problems have been fully resolved and the staff is fully f amiliar with the operating characteristics of the new core configuration, there should be no change in operating procedures which might result in shorter response time or more seriuus radioactive contamination in the event of a LOCA.
Experience has shown that human error can often compound a relatively simple emergency, resulting in consequences much more serious than would have been anticipated. .
(b) The mathematical analysis performed by the Licensee assumes a core of either LOPAR or OFA fuel. Since the l ,[ bpomogeneousfulltransitional mixed core has not been studied as such, it would
,p he wise to wait until after the transition is complete before
[h translating the results of such analysis into licensing changes.
c jThe BART model used in the Licensee's analysis assumes (for N\ Jpurposes of mathematical convenience) that the system pressure b tl/ fis constant. In doing so, a number of phenomena which could kJ p g significantly increase the cladding temperature have been excluded from consideration, including flow stagnation and flow hyg oscillation. Moreover, the BART model does not encompass all ybp[ possible expected flow patterns even if he system pressure is g k i relatively constant. % cggg hig Qg M W,hd
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(/h54 W,W 4 h (d) The BART model does not include a gap heat transfer model or a and such phenomena as embrittlenent, O j[ cladding blistering,swelling hydriding,model, swelling, bowing, fission gas pressure,
- 3 and possible reactions which might permanently affect heat N transfer rates are not featured in the tests (involving D,Dr.d electrically heated fuel arrays) that were used to ascertain the U validity of the BART predictions. Such phenomena, by weakening certain localized parts of the fuel cladding, or by causing M q () 50 t - e kedes ce&.s
hw4 gt.akus M Vv-C l LM e u d undercoo ng in cert ai smal J;k % region ( , c eDU d materially affect the probability of cladding failure (since "a chain is only as strong as its weakest link"). To cope with such important
( problems simply by assigning a numerical " penalty" is little more than blind gue swork, reflecting the limitati of analysis. fMg AMtva._ M QCQ (e) According to the analysis based on the FLECHT correlation, clad temperature could reach 2140*F: Just 60*F short of the limit, which is 22OO*F. Given the approximate nature of the analysis, it is purely a matter of Judgment -- political, rather than technical in nature - as to whether or not this is "too close '
for comfort", especially since it is known that the reflood will l be slowed down by the increased hydraulic resistance of the
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mixed core and that the shut-off rods will take more time to drop e s g]q (g g+q q gQhgg (f) A computer analysis involving a great many calculations does not generally produce results which are perfectly accurate. Errors
.O of various kinds creep into the results, some of which are easy 4
p / to understand -- suchTheasEconomic round-off error of Council -- and others Canada which are has reported y i Y] much more subtle.
that its CANDIDE model for the Canadian economy sometimes Mo ' f. develops explosive cycles for no apparent reason, yielding results in which the error is very much larger than the answer.
N sfAtpresent, there is no way of predicting in advance the maximum L error t' hat may obtaan in any given comput W Efal'ysis. This is an p4 pepp k over-rfatng difficuli.7 ,,i i.h al l coalsiuter analysis invo Wing large numbers of calculations. ^
i k h(g) In it b fety Evaluation Report on the BART Code, referenced by YkM Cecil D. Thomas of the Licensing Division on December 21~1983, k we learn that "The BART code shows spikes in the calculated h ,b results of the heat transfer coefficients. The spikes are indicative of the discontinuous heat transfer regime A( However, the overall BART predictions are in good c' agreement with the heat transfer coefficient data." Of course,
%%a^[f% transitions.
g it is precisely where the " discontinuous heat transfer regime c Sg transitions" occur that cladding failure is likely to occur.
- g Cladding failure generally begins as a local phenomenon, not 3 t/ necessarily as an "overall" phenomenon. ,
(h) From the same NRC document, we learn that "the analyses show
% that BART consistently overpredicts the average clad temperature." Again, it is not the average clad temperature that determined whether or not cladding failure will occur, but the localized clad temperature. In the vicinity of a discontinuity, an acceptably low average value does not translate into an acceptably low localized value.
(i) To prevent DNB, the NRC has employed the concept of a local DNB ,
heat flux ratio, defined as the ratio of the critical heat flux (that would cause DNB at a particular core location) to the actual local heat flux: DNBR = CHF/AHF. If DNBR is less than or equal to 1, then DNB will occur. However, because of the R
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e many uncertainties that exist, even-if DNBR is calculated to be greater than 1, DNB may st,ill occur (because the actual DNBR is g
not the same as the calculated DNBR). The DNB design basis is that there must be at least a 95 percent probability, with 95 percent confidence, that DNB will not occur provided that the calculated DNBR is greater than a certain specified limit. !
Thus, even if the calculated DNBR does not fall below the specified limit, there is still a chance that DNB may occur.
Under the circumstance, if the reactor is operated in such a way that the fuel runs at a hotter temperature, even if the calculated DNBR is kept above the prescribed limit, the probability of DND (resulting in fuel cladding failure) is materially increased.
(j) The Licensee argues that lowering the DNBR limit f rom 1.3 to l i
1.17, "in no way implies a reduction in the safety margin of a nuclear reactor" [ Licensee's Motion for Summary Disposition of ;
Intervenor's Contention (d)]. This statement is incorrect. By allowing the fuel to run at a hotter temperature, the reduction in DNBR limit does allow for a greater probability that DNB will occur. What the Licensee intends to say, perhaps, is that re-calculation indicated that the same margin of safety that was previously thought to exist can now be achieved at a higher operating temperature. Whatever the truth of this latter statement, it is undoubtedly true that running at a hotter temperature materially increases the probability of DNB, and therefore reduces the safety margin of the nuclear reactor.
(k) Given the limitations of mathematical analysis, all of these conclusions should be carefully tested against the actual operating experience of the plant. If the analysis is correct,
- then there should be no history of fuel failures and therefore l no iodine releases.
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- 10. Bo Lindell, Chairman of the International Commission for Radiological Protection, has commented on the oft-quoted dictum that all radiation doses should be kept "as low as reasonably achievable".
He writes, "On the ALARA principle, doses near the dose limit would be rare, and the limit must now be seen as indicating the region of undisputed unacceptability rather than as guidance on what might still be acceptable" C" Basic Concepts and Assumptions Behind the New ICRP Recommendations", IAEA-SR-36/51, March 19793. Since DNB or inadequate reflooding following a LOCA will result in releases of radioactivity, this is an appropriate philosophy to adopt in relation to reactor 8
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safety as well. Contrast the Licensee's attitudinal stance, that running the fuel at a hotter teinperature, provided it is within certain specified limits, "in no way implies a reduction in the safety margin of a nuclear reactor." Such a statement is not only technically incorrect, but inconsistent with a genuine concern for maintaining reactor risks "as low as reasonably achievable".
SWORN BEFORE BE AND I HAVE SIGNED.
at Saint-Laurent, Province of Quebec g, 73 this 30th day of August 1984. " @
MTRE ADRIDI LEDUC, flotary 9
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EXHIBIT A Professional Qualifications and Experience of Gordon D. J. Edwards My name is Gordon D. J. Edwards and my home address is 1300 Raimbault, Ville St. Laurent, Quebec, Canada, H4L 4R9. I am a professor of Mathematics and Science at Vanier College in Montreal, and I am also President of the Canadian Coalition for Nuclear Responsibility (Inc).
I graduated from the University of Toronto with a Bachelor of Science degree in Mathematics, Physics and Chemist v (gold. medal in mathema-tics and physics) in June, 1961. As a Woodrow Wilson Fellow, I graduated from the University of Chicago with a Master of Science in Mathematics in June, 1962, and with a Master of Arts in English Language and Literature in June, 1964. After teaching university-level mathematics at the University of Western Ontario for several years, I graduated from Queen's University (Kingston, Ontario) with a Doctor of Philosophy degree in mathematics in June 1972. I conducted i
post-doctoral research in the Economics of Ocean Fisheries at the University of British Columbia in 1972-1973, and was the Assistant Director of a nation-wide study of the Mathematical Sciences in Canada j for the Science Council of Canada in 1973-1974.
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i As an applied mathematician, I have been keenly interested in the strengths and weaknesses of mathematical modelling as it is applied to t'
real-life problems. In 1977, I was a consultant to the Cluff Lake i
Board of Inquiry into Uranium Mining in Saskatchewan, where I cross-examined technical experts on such subjects as health effects of 1
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radiation, radioactive waste disposal, and reactor safety. In 1977-( 1978, I was a consultant to the Ontario Royal Commission on Electric Power Planning, where I cross-examined experts in reactor safety from Ontario Hydro (the utility), Atomic Energy of Canada Limited (a research and development organization), and the Atomic Energy Control Board (the regulatory agency) over a period of several months. In 1979-1980, I was a consultant to the Select Committee on Ontario Hydro Affairs (a Committee of the Ontario Legislature) during a thirteen-week investigation into reactor saf ety f ollowing the Three Mile Island accident. I have also acted as a consultant on nuclear-related matters to such bodies as the Canadian Broadcasting Corporation, the National Film Board, the Science Council of Canada, the United Steelworkers of America, and several Canadian governmental bodies.
I have published several articles on nuclear power in Canada, with special reference to reactor safety and economics; in particular,
" Cost Disadvantages of Expanding the Nuclear Power Industry" (Canadian Business Review, Spring 1982) and " Canada's Nuclear Dilemma" (Journal I of Business Administration, Vol.13 1982). I have also prepared l
numerous unpublished reports involving technical critiques of various safety analyses for various clients.
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