ML20203N112
| ML20203N112 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs, 05000000 |
| Issue date: | 01/06/1986 |
| From: | Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20198G688 | List:
|
| References | |
| NUDOCS 8609230147 | |
| Download: ML20203N112 (2) | |
Text
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Ed, MEMORANDUM FOR: Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM: Thomas E. Murley, Regional Administrator Region I -
SUBJECT:
COMMENTS ON CALVERT CLIFFS EMERGENCY PLANNING ZONE (EPZ) EXEMPTION REQUEST
, The Region I staff has reviewed the Calvert Cliffs application for exemption from the generally accepted 10-mile EPZ requirement and provides the following comments.
It is generally agreed that the generic 10-mile EPZ may not necessarily represent the actual emergency planning needs of each operating reactor site and that an EPZ of other than a 10-mile radius may be deemed valid after consideration of onsite and offsite variables. However,,the consideration of, this exemption request should be approached cautiously. Two important questions must be answered before this request, and others of this type which may follow, can be considered solely for their logistical and technical merits. The first question is . . . Who determines the size of the EPZ? BG&E states in the application that only those persons most likely to be subjected to emergency conditions would be located in the EPT and concludes that public
-health would be protected. One may argue that the state and local governments should play a major role in determining who is to be protected and subsequently in setting the size.of the EPZ. The State of California, for example, requires planning for an Expanded Planning Zone (20-mile radius). Th'ere is very little mention in the Calvert Cliffs application of how the state and local govern-ments are to be involved in the decisionmaking process.
The second question is . . . Should the consideration of this exemption request parallel the pace of present generic source term studies or should it be considered on a plant-specific basis? The consensus is that while certain aspects of such an application may be considered generically (source term),
some aspects must be considered on a plant-specific basis (plant system configurations (Event V), containment design, EPZ demographics, . . .). The most conservative approach would be to consider and resolve the generic questions prior to considering exemptions on a plant-specific basis. This is our recommendation.
The following comments are also provided: )
- 1. The studies supporting the exemption request were not attached.
Consideration of this documentation must obviously be included as a '
part of the review process.
- 2. There wis no indication made in the Calvert Cliffs application of the planning and planning zone considered for the ingestion pathway, 8609230147 860910 PDR COMMS NRCC --
CORRESPONDENCE PDR
Harold R'. Denton -
- 3. .The BG&E' submittal appears to assume the availability of the Containment Spray System for~ source term consideration.
Please contact Mr. Terry Harpster of my staff at FTS 488-1208 if ~you have -
questions regarding these comments.
. . Thomas E. Murley Regional Administrator cc: V. Stello, DEOROGR '
J. M. Taylor,- IE
-G.-H. Cunningham, ELD a
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- .. b JAN 0 31986 The Honorable Roy Dyson United States House of Representatives Washington, D.C. 20515
Dear Congressman Dyson:
Thank you for your letter of December 6,1985 to Chairman Palladino concerning the November 18, 1985 application from Baltimore Gas and Electric Company (BG&E) to reduce the size of the emergency planning zone (EPZ) for Calvert Cliffs Units I and 2. I am writing to advise you of the status of the staff's review of BG&E's request.
BG&E's application is for an exemption from the Comission's regulations which require that, generally, the plume exposure pathway EPZ for nuclear power plants shall consist of an area about 10 miles in radius. The NRC staff is presently in the process of identifying the legal and technical issues associated with BG&E's request. BG&E's application will be carefully
, evaluated in light of the present status of ongoing analyses regarding
_ source term data and methodology and reference plant risk profiles. At this time, the staff's review has not progressed to the point where it could address the merits of BG&E's request. Let me assure you, however, that the -
staff is sensitive to the public's interest in the issues raised by BG&E's request and that any proposed action by the staff will be in full considera-tion of that public interest.
Thank you for your coments concerning BG&E's application to reduce the size of the EPZ for Calvert Cliffs. If we can be of further assistance, please contact us.
Sincerely, OM -*.0. c' r ed by J a c. .. . o William J. Dircks Executive Director for Operations 7(aoll o 4c4 7 l
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CONGRESSMAN ROY DYSON 224 CaseNoN Houst OFFict Butto*NG. WasMmCTON. D.C. 20615 (2021225 5311
' December 6, 1985 s
b Mr. Nunzio J. Palladino
- . Chairman
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Nuclear Regulatory Commission - -
1717 H Street, N.W. '
Washington, D.C. 20555
Dear Mr. Palladino:
I write regarding Baltimore Gas and Electric Company's proposal to reduce the emergency planning zone (EPZ) surrounding its Calvert i Cliffs nuclear generating facility in Calvert County, Maryland. After careful consideration of this issue and discussions with elected L i
officials-and area residents, I must oppose this proposal. There is no '
- i .
justifiable reason to reduce the EPZ from ten miles to two miles.
f The residents of the area immediately adjacent to, and surrounding, a nuclear power plant deserve the full attention of the utility operating the plant and the most comprehensive plan available .
j to protect them from the effects of an accident. An EPZ extending ten miles from a plant represents the traditional level of protection - ,
l required by the Commission, and is not unreasonable given the dangers associated with a nuclear plant. There is little scientific evidence
~to demonstrate precisely how much damage a nuclear accident could inflict on people and the environment near the site. In the absence ;
of such evidence, the reasons supporting the original establishment of.
the EPZ at ten miles remain, and that size represents the minimum protection the inhabitants of the area should have.
i The ten-mile zone surrounding the Calvert Cliffs facility has L been in place for several years. Baltimore Gas and Electric (BG & E)
I has advanced no valid reason to reduce it. The company and the local ;
I governments have developed and implemented the necessary evacuation plans and communication plans to be used during an emergency.
Residents and concerned citizens have also planned for an emergency based on the existence of the ten-mile zone. These plans were developed through substantial investment of time and money by all i
parties involved, and BG & E admits that this reduction will not substantially reduce.its customers' bills.
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Mr. Nunzio J. Palladino December 16, 1985
--page 2--
BG & E believes the EPZ should be reduced to reflect more accurately the actual danger associated with an accident, and to symbolize to the residents of the area that Calvert Cliffs is safe.
BG & E points to both the safety record of the plant and to recent scientific evidence indicating that the danger of serious accidents
. may be lower than earlier believed.
Fortunately, the nuclear power industry has not experienced a disaster approaching the worst case scenario. But, to reduce the size
- of the EPZ purely as a symbolic gesture is both irresponsible and ,
unsound public policy. Until the company can prove that a two-mile n area encompasses the total area likely to be affected by any nuclear accident, I must insist on the current ten-mile EPZ.
i Until BG & E can prove that the EPZ can be reduced without ri k ;
- to those in the present ten-mile EPZ, I cannot support any reduction.
I strongly urge you to deny the request of BG & E to reduce the size i
of their EPZ. It is our responsibility to ensure that this minimum -
j protection is not altered.
Thank you for your attention in this very important matter. If you require additional information or if I may be of further -
assistance, please do not hesitate to contact me.
e With kindest regards, I am g4.---,.. .
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j ROY SON ,
.MEMB OF C NGRESS ;
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A S ALTIMO RE GAS AND ELECTRIC CHARLES CENTER
- R O. BOX 1475 BALTIMORE, MARYLAND 21203 JoS EPH A.TIERNAN VICE PRESACENT NUCLEAR ENERGY 3anuary 10, 1986
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U. S. Nuclear Regulatory Commission '4 Office of Nuclear Reactor Regulation O RECEfyED Washington, D. C. 20555 _ q-ATTENTION: Mr. Ashok C. Thadani, Director W amERarraummg1 PWR Project Directorate #3 -
P Division of PWR Licensing-B 9 SUB3ECT: Calvert Cliffs Nuclear Power Plant Q A t.P Unit Nos.1 & 2; Docket Nos. 50-317 & $0-313 Request' for Amendment
REFERENCE:
(a) BG&E letter from Mr. 3. A. Tiernan to Mr. E. 3. Butcher, Jr, dated November 18,1985 Gentlemen:
In Reference (a) Baltimore Gas and Electric Company requested an exemption from the requirements of 10 CFR 50.47(c)(2) and 10 CFR 50 Appendix E to the extent that the plume exposure pathway emergency planning zone (EPZ) surrounding the Calvert Cliffs Nuclear Power Plant would be redefined as the area within a radius of two miles from the plant instead of the generally required radius of about tea miles.
Our request was filed under the provisions of 10 CFR 50.12 which establishes certain standards for evaluating the merits of an exemption request, but which provides little guidance to the licensee or the NRC regarding the mechanism and process for staff review and public participation.
Upon consideration of these additional issues, we have determined that this exemption request should be also processed as a license amendment in that the exemption would I constitute a significant change in the emergency plan. 10 CFR 50.54(q), which is referenced in License Condition ll.C of the Unit I and Unit 2 Operating Licenses, requires that any changes that may reduce the effectiveness of the emergency plan be submitted as an application for NRC review and approval. Our understanding is that such ,
approval must be sought via the provisions of 10 CFR 50.90 to the extent that a decrease '
in the effectiveness of the plan could be construed to be an unreviewed safety question or to involve significant hazards considerations since the consequences of accidents described in the FSAR might be increased.
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Mr. Ashok C. Thadani January 10,1936 Page 2 Therefore, pursuant to 10 CFR 50.90, Baltimore Gas and Electric Company hereby requests Amendments to its Operating Licenses No. DRP-53 for Unit I and No. OPR-69 for Unit 2 which would allow the proposed change described in Reference (a) in the
- context of a possible significant hazards consideration.
SAFETY COMMITTEE REVIEW This proposed change has been reviewed by our Plant Operations and Off-Site Safety-Review Committees, and they have concluded that implementation of this change will not result in an undue risk to the health and safety of the public.
FEE DETERMINATION Pursuant to 10 CFR 170.21 we are including BG&E Check No. (/M24) in the amount of -
$150.00 to the NRC to cover the application fee for this request.
Very truly yours, A
I UWW J. A. Tiernan Vice President - Nuclear Energy ,
STATE OF MARYLAND :
- TO WIT:
CITY OF BALTIMORE :
Joseph A. Tiernan, being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belief; and that he was authorized to provide the response on behalf of said Corporation.
WITNESS my Hand and Notarial Seal: 4 Lc .t Notary Public /
My Commission Expires: 7 fb
' Date
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JAT/BSM/ dim Attachments
1
, t Mr. Ashok C. Thadani
-January 10, 1986 Page 3 cc: D. A. Brune, Esquire
- 3. E. Silberg, Esquire D. H. Jaffe, NRC T. Foley, NRC T. Magette, DNR 0
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-! t Mr. Ashok C. Thadani
' January.10,'1936 Page 4 -
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bcc: R. F. Ash /R. C. L.- Olson C. H. Cruse /P. E. Katz R. E. Denton/3. A. Mihalcik R.' M. Douglass/T. N. Pritchett Dr. M. Gavrilas/E. I. Bauereis
- 3. R. Lemons /R. P. Heibel W. 3. Lippold/A. R. Thornton F. 3. Munno R. B. Pond /R. E. Cantrell L. B. Russell /3. T. Carroll V. F. Stricklin (3)
B. S. Montgomery -
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M. 3. Warren M. E. Bowman L. E. Saiyards s
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CHARLES CENTER.P. O. BOX 1475 8ALTIMCRE. MARYLAND 21203 ARTHUR c. LUNOVALL. JR.
VtCC PRESiOEse?
Supptv November 13,1935 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission .
Washington, DC 20555 Attn: Mr. E. 3. Butcher, Jr., Chief Operating Reactors Branch //3 Division of Licensing
Subject:
Calvert Cliffs Nuclear Power Plant Units ! and 2; Doc!<ets 50-317 and 50-313 Recuest for Exemotion -
Gentlemen:
Pursuant to '10 CFR 50.12(a), Baltimore Gas and Electric Company hereby requests an exemption for Calvert Cliffs Nuclear Power Plant (CCNPP) Units Nos. I and 2, Facility Operating Licenses DPR-53 and -69, respectively, from a requirement of 10 CFR
. 50.47(cX2) and 10 CFR 50 Ap?endix E, as described below.
Exemotion Recuest In accordance with the provisions of 10 CFR 50.12(a), we request an exemption from the l general requirement of 10 CFR 50.47(cX2) and Section I of 10 CFR 50 Appendix E, which states that the plume exposure pathway emergency olanning zone (EPZ) should consist of an area of about 10 miles in radius. Specifically, we request that the plume exposure pathway EPZ surrounding CCNPP be redefined as the area within a radius of 2 miles l from the center of the containment / auxiliary building complex, based on our l determination that a plume exposure pathway EPZ beyond 2 miles is not necessary to achieve the underlying purpose of the rule, as discussed in Attachment 1, and that there j now exist significant material circumstances which were not available for consideration j
when the regulation in question (i.e., the 10 mile radius) was adopted, as discussed in ,
Attachment 2. !
Our determination relative to the adequacy of the proposed 2-mile EPZ was achieved using the same regulatory philosophy and basic aporoach as that presented in NUREG- 1 0396, but utilizing current source term information. !
Te have concluded that the requested exemption is authorized by law, will not endanger life or property or the common defense and security and is otherwise in the public interest. -
The exemption is in the public interest for a number of reasons. The size proposed for our plume exposure pathway EPZ is based upon a more recent, more rigorous and more g @ 8 9 \\ 4 9 l pri
t o.. -c Mr. E. 3. Butcher, Jr. Noverrber 13, 1935 accurate body of knowledge than that which formed the basis for the existing EPZ size, as discussed in Attachment 2. Consequently, the more realistic EPZ size would allow-BG&E and the applicabic government agencies to focus and acply their emergency planning resources more efficiently within the area that is most likely to be subject to actual emergency conditions. The exemption would also eliminate the unnecessary regulatory burden of emergency planning from those local jurisdictions which would no longer be included in the plume exposure pathway EPZ. As currently configured, the EPZ '
encompasses portions of three counties: Calvert, Dorchester and St. Mary's. The proposed EPZ .would lie completely within 'Calvert County. The rationale for the conclusion that life and proper,ty will not be endangered is inherent in the basis for the proposed reduced EPZ size, as discussed in Attachments 1 and 2.
Safety Committee Review This request has been reviewed by our Plant Operations and Safety Review Committee and Off-Site Safety Review Committee, and they have concluded that, if granted, it will l not result in an undue risk to the public health and safety.
Fee In accordance with 10 CFR 170.12 and 170.21, a check is enclosed in the amount of
$150.00 as payment for the fee for this application.
Very truly yours,
- bh for A. E. Lundvall, Jr. ~N Vice President, Supply STATE OF MARYLAND :
- TO WIT:
CITY OF BALTIMORE : .
Joseph A. Tiernan, being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he. provides the foregoing. response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information and belief; and tipt ,he was authorized to provide the respense on behalf of said Corporation. I
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WITNESS my Hand and Notarial Seal: Ibe'I g ,
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Attachments: ~ ;
- 1. The Rule and Its Underlying Purpose. -
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- 2. Material Circumstances Not Considered When the Regulation in Question was Adopted.
cc: D. H. Jaffe Dr. T. E. Murley D. A. Brune, Esq.
G. F. Trowbridge, Esq.
1
.o' P. t <.- Air. E. 3._Butchtr, Jr. - Noverrber .13,193 5 -
bec: Messrs: B. C. Trueschler .
. G. V. McGowan '
'A. E. Lundvall, Jr.-
- 3. A. Tiernan C. H. Poindexter T.~ E. Forgette
.R. C. L. Olson S. M. Mirsky T. L. Cook R. N. M. Hunt B. S. Montgomery J. P. Bennett G. W. Gephart C. 3. Franklin E. 3. Neumann R. F. Ash /A.' R. Thornton G. C. Creel /3. R. Lemons /D. L. Pobloskie C. H. Cruse'/T. N. Pritchett R. M. Douglass/A. B. Anuje
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Dr. M. Gavrilas/Or. E. I. Bauereis .
W. 3. Lippold/M. 3. Miernicki R. E. Denton/M. E. Bowman Dr. R. B. Pond, Jr./L. E. Titland L. B. Russell /3. T. Carroll
- Dr. F. 3. Munno -
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ATTACHMENT 1 THE RULE AND ITS UNDERLYING PURPOSE
- 1. Introduction This attachment discusses our determination that a plume exoosure pgthway EPZ beyond' 2 miles is not necessary to achieve the underlying purpose of the rule as applied to the
' Calvert Cliffs Nuclear Power Plant (CCNPP).
10CFR50.47(c)(2) states, in part:
" Generally, the o!ume exposure pathway EPZ for nuclear power olants shall consist
, of an area about 10 miles (16km) in radius..."
Footnote I to :10CFR50- Appendix E cites NUREG-03962 .for a discussion of EPZs for power reactors. NUREG-0396 was prepared by a joint U.S.' Nuclear Regulatory' Commission and . U.S. Environmental Protection.- Agency Task Force on Emergency Planning. As the planning basis document for emergency planning, it. represents the ,
primary published source of understanding of the underlying purpose of the existing rule.
Specifically, NUREG-0396 (page 24) states:
"The estab!ishment of Emergency' Planning Zones of about 10 miles for the plurne exposure pathway . . . is sufficient to scope the areas in which planning for the initiation of predetermined protective action is warranted for any given nuclear plant."
The following sections summarize (a) the underiving purpose for the establishment of'the.
plume exposure pathway EPZ and the basis for 10 miles as an appropriate EPZ radius, as discussed in NUREG-0396, and (b) the achievement of the underiving purpose of the subject rule in the particular circumstances of CCNPP.
- 2. The Underiving Purocse of the Plume Exoosure Pathway EPZ and the Basis for 10 Miles Insight into the underlying purpose of the rule establishing a plume exposure pathway EPZ radius (of about 10 miles) may be gained by referring to the following statement on page 5 of NUREG-0396:
"The Task Force concluded 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."
Dose savings would be achieved by protective actions within the plume exposure oathway EPZ, so that the doses in question would not be received by individuals. Therefore, the underlying purpose of this plume exposure pathway EPZ is to establish an area within i
I 10CFR50.47(c)(2) and Appendix E.
2
" Planning Basis for the Develooment of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants", ,
, NUREG-0396 (EPA-520ll-73-016), U.S. Nuclear Regulatory Commission, December 1978.
, s' 9 ~s which planning of predetermined orotective actions would be exoe.cted to result in dose savings comparable to those discussed in NUREG-0396.
The following quotations from NUREG-0396 further . address the conclusions and )
recommendations of the Task Force with regard to the basis for the establishment of a planning radius for the plume exposure pathway EPZ:
In discussing the r.eed for planning zones, NUREG-0396 on page 3 and 11, respectively, states:
"The most important guidance for planning officials is the distance from the nuclear facility which defines the area over which planning for pre-determined actions should be carried out. The other elements of guidance provide succorting information for planning and preparedness."
"With regard _to the area over which planning efforts should be carried out, the Task Force . recommends that ' Emergency Planning Zones' (EPZs) about each nuclear facility be defined both for the short term ' plume exposure pathwav' and for the longer term ' ingestion exposure pathways."' .
With regard to consideration of ootential accidents which might result in the above mentioned exposures, NUREG-0396 states, on page 24 and 15, respectively:
"A spectrum of accidents (not the source term from a single accident sequence) should be considered in develooing a basis for emergency planning."
"The Task Force agreed that emergency response plans should be useful for responding to any accident that would produce offsite doses in excess of the PAGs.
This would include the more severe design basis accidents and the accident spectrum analyzed in the RSS. Af ter reviewing the potential consequences associated with these types of accidents, it was the concensus (sic) of the Task Force 'that emergency plans could be based upon a generic distance out to which predetermineel actions would provide dose savings for any such accidents. Beyond this generic distance it was concluded that actions could be taken on an ad hoc basis using the same considerations that w
- Protective Action Guides;gntRSS -into the initial Reactor Safety action Study .)detgrminations." (NOTE: PAGs In reaching the conclusion on cage 24 that a olume exposure oathway EPZ "of about 10 miles. . .is sufficient", NUREG-0396 cites technical data contained in Appendix I to that report, entitled " Rationale for the Planning Basis". Apoendix I includes a discussion of various rationales for establishing a planning basis.
The consequences of Design Basis Accidents (DBAs) were calculated for a large number of nuclear oower plants. Those calculations were related to the PAGs, i.e.. Figure I-8 of Appendix I to NUREG-0396.
3
" Manual of Protective Action Guides and Protective Actions for Nuclear Incidents",
EP A-321/1-75-001, U.S. Environmental Protection Agency, September 1975.
'4 Reactor Safety Study "An Assessment of Accident Risks in U.S. Commercial
, Nuclear Power Plants". WASH-1400 (NUREG-75/1014), U.S. Nuclear Regulatory Commission, October 1975. .
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3 w It should be noted that. NUREG-0396 expressed explicit caution in the use of PAGs in establishing requirements for emergency response plans. On page 4, NUREG-0396 states:
"The nature of PAGs is such that they cannot be used to assure that a given level of exposure to individuals in the population is prevented. In any particular response situation, a range of doses may be experienced, principally deoending on the distance from the point of release. Some of these doses may be well in excess of the PAG levels and clearly warrant the initiation of any feasible protective actions. This does not mean, however, that doses above PAC levels can be orevented or that emergency resoonse olans should have as their 05iective oreventing doses above PAG levels."(emphasis added)
Figure I-11 of Appendix I to NUREG-0396 was a orimarv figure in the develooment of the rationale for recommending a planning basis incoroorating a 10 mile radius for the plume expos'ure pathway EPZ. The data in that figure were based on the then-current, best available data from the Reactor Safety Study.
Taken together, the data presented in Appendix I of NUREG-0396 and the results of the .,.
investigation summarized therein formed the basis for the establishment of the plume exposure pathway EPZ with a radius of about 10 miles.
,3 . Achieving the Underivine Purcose of the Rule As noted in NUREG-0396, the principal exoosure sources in the plume exposure oathway
, are: (a) whole body external exposure -to gamma radiation from the plume and from t
deposited material: and (b) inhalation exposure from the passing radioactive plume.
Analyses of these exposure sources in the plume exposure pathway have been ocrformed for CCNPP covering a spectrum of postulated accidents, as presented in Figures 1-1 and
- . 1-2. These analyses include the Design Basi Safety Analysis Report (FSAR) for CCNPP,g and theAccidents (D83A),
more severe asaccidents.
" Class 9" reoorted in the F The latter are based on uodated source terms (i.e., the magnitude, type and timing of postulated releases of radionuclides to the environment), as oresented in Attachment 2, in combination with the Calvert Cliffs site-specific meteorological data, as repogted in the FSAR. These latter analyses were performed with the CRAC2 computer code . The CRAC2 analyses assumed no evacuation and 24-hour normal activity (i.e., no special sheltering.) '
4 The solid curve in Figure 1-1 deoicts the whole body dose as a function of distance from the point of release for the DBA at CCNPP.' This curve was reproduced from Figure 14.24-2 of the FSAR. It can be observed from this curve that the 5 rem PAG is not exceeded beyond 0.4 mile and the I rem PAG is not exceeded bevond 1.4 miles. s Whole body doses as a function of distance, as reported in Figure I-S of NUREG-0396, are 5
" Updated Final Safety Analysis Report - Calvert Cliffs Nuclear Power Plants - Units 1 and 2", Baltimore Gas and Electric Company, Dockets 50-317 and 50-313. ,
6 "CRAC2: Calculations of Nuclear Reactor Accident Consequences, Version 2, NUREG/CR-2326, Sandia National Laboratories, February 193*3.
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if j included here as ' the dashed and dotted curves in Figure 1-1, reoresenting the . median
,(50%) and highest'10% of calculations for 67 cower olant sites considered in NUREG-0396.-
F om Figure 1-1,'it can be seen that a comoarable dose, e.g. 0.5 rem, is projected at a
- d. stance of 2 miles for CCNPP and at a distance of 10 miles for the highest 10% curve obtained from NUREG-0396. Thus, dose _ savings with a 2 mile EPZ at CCNPP would be
. comparable to dose savings with a 10 mile EPZ, based on the NUREG-0396 data.,
Figure 1-2 presents data deoicting the probability of experiencing a 200 rem whole body dose, at which significant,early injuries may start to occur, as a function of distance from the point of release for the (more severe) Class 9 accidents. This figure includes an adaptation of the data presented in Figure I-Il of NUREG-0396 (i.e., the dotted curve in
' Figure 1-2) and two curves deoicting the data for a range of conditions at CCNPP.
The dotted curve comprises the corresponding curve in Figure I-11 of NUREG-0396, as adapted for. Figure 1-2 by multialving the conditional orobabilities (conditional on a$ core melt accident _ occurring) times the WASH-1400 core melt probability of 5 x 10' per reactor year, reported in Figure I-il. -
Two CCNPP site-soecific curves are included in Figure 1-2, based on the recent source terms presented in - Attachment 2. Both curves include contributions from the' in-containment accident sequences. The difference between the two curves is due to thb
'different contributions from the two containment bypass "V" sequences.
As discussed in Attachment 2, " low range" and "high range" V sequence source terms have been established for CCNPP. When the high range V sequence. source term is included along with the in-containment source terms, the data depicted in the "V-High" curve of Figure 1-2 result. When the low range V sequence source term is included along with the in-containment source terms, the data depicted in the "V-Low" curve result.
Thus, these two curves depict the effect of the range of potential V sequences on the probability of exoeriencing a 200 rem dose as a function of distance. The fact that the two curves arc *close together is an lilustration of the containment'so Jrce terms obtained from NUREG-0956.jarge
~
In thecontributions frominthe cases presented thisin-figure, the probability of experiencing a 200 rem dose at any distance is substantially lower for CCNPP, with the current source terms, than was the case in the NUREG-0396 analysis, with WASH-1400 source terms.
The CCNPP dose curves are observed to drop off substantially at a distance of about 2 miles, whereas the NUREG-0396 dose curve is observed to drop off substantially at about 10 miles. Thus, dose savings with a 2 mile EPZ at CCNPP, based on recent source terms,
! are comparable to dose savings with a 10 mile EPZ, based on WASH-1400 source terms as reported in NUREG-0396.
Taken together, the data presented in Figure 1-1 and 1-2 illustrate that the underiving .
purpose of 'the EPZ plume exposure pathway rule, as discussed above, is achieved at a ,
p radius of 2 miles at the Calvert Cliffs Nuclear Power Plant. .q 7
Silberberg, M. et. al., " Reassessment of the Technical Bases for Estimating Source Terms", NUREG-0956 Draft Report for Comment, U.S. Nuclear Regulatorv Commission, July 1935.
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CALVERT CLIFFS and NUREG.-0)96 ' Colculation 2 hr. Whole Body Dose for Licensing Calculation of DBA/ LOCA at 2 Hours Assuming 5 Percentile Meteorology and Straight Line Plume Trajectory
( Reproduced from Fig.1-8, NUREG. -0396)
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ATTACHt1ENT 2 MATERIAL CIRCUMSTANCES NOT CONSIDERED
~ WHEN THE REGULATION IN OUESTION WAS ADOPTED 1.0 Introduction The requirement of 10CFR 50.47 for a p!qme exposure pathway Emergepcy Planning Zone (EPZ) with a radius of about 10 miles is discussed in NUREG-0396, as noted in footnote 1 to 10CFR50 Appendix E, Section I.3 The technical bases .for the recommendations contained in NUREG-0396 are predicated on a revieiv of a spectrum of potential accidents ranging from the Design Basis Accidents (DBAs) to' the more severe " Class 9" accidents. The analysis of the Class 9 accidents in NUREG-0396 is based on source terms (i.e., the magnitude, type, and timing of postulated releases of radionuclides to the environment) as reported in the Reactor Safety Study (WASH-1400),, which reoresented the best technical information available at the time.
There are now present other material circumstances not available for consideration when -
the regulation was adopted based on the WASH-1400 data. Specifically, vast improvement in the knowledge and understanding of source terms for postulated severe core damage . accidents has occurred as a result of the extensive national and international reassessment of severe accident source terms. This attachment addresses-the recent improvements in source term knowledge and the application of that knowledge to the Calvert Cliffs Nuclear Power Plant (CCNPP).
2.0 Recent Imorovements in Source Term Knowledge
2.1 Background
WASH-1400 addressed a broad range of potential accident sequences -and assigned probabilities of occurrence to each sequence. It also addressed the release of radionuclides for the various sequences and grouped the secuences into release categories. Nine release categories were developed for Pressurized Water Reactors (PWRs) and five release categories were developed for Boiling Water Reactors (BWRs).
For each..of these release cate;;ories, the authors of WASH-1400 established source 1
" Planning Basis for the Development of State and Local Government Radiological Emergency Resoonse Plans in Suoport of 1.ight Water Reactor Nuclear Power Plants", NUREG-0396 (EPA-520ll-73-016), U.S. Nuclear Regulatory Commission, December 1978.
9 1
3 10CFR50 Appendix E, " Emergency Planning and Dreparedness for Production and Utilization Facilities".
4
" Reactor Safety Study - An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants", TASH-1400 (NUREG-75/104), U.S. Nuclear Regulatorv Commission, October 1975.
. o 9
terms, which included estimated releases to the environment within seven fission product groups. These seven grouos can be categorized into three categories of releases - noble gases (xenon and krypton), volatile species (iodine, cesium-rubidium, and the tellurium-antimony groups), and non-volatile species (barium-strontium, ruthenium, and the 1anthanum groups.) For the purposes of this exemption request, only the PTR sequences and releases are specifically discussed.
The first of these categories of releases, noble gases, is not influenced appreciably by plant features and physical phenomena, except for the timing of release. A delav in the time of releasa affects the noble gas source term because it allows time for natural radioactive decay to reduce the amount of radioactivity available for release to the environment. For accident sequences involving postulated early releases to the environment, the WASH-1400 investigations assumed essentially complete release of the noble gases to the environment. The recent assessment of severe accident source terms has not altered that assumption. With respect to the two other categories of releases, volatiles and non-volatiles, the new knowledge and improved understanding resulting from the extensive investigation of severe accident source terms has resulted in substantive reductions in the estimated fracticn of the core inventory of these radionuclides which would be expected to be released to the environment (b, substantially reduced source terms when compared with WASH-1400, as discussed in subsequent sections of this attachment.)
One early finding concerning imorovement in the methodology of estimating releases g the environment during severe accidents was reported in an August 4,1980 letter addressed to the then Chairman of the Nuclear Regulatory Commission, 3. Ahearne, from three respected scientists, W. R. Stratton of the Los A!amos Scientific Laboratory, and A. P. Malinauskas and D. O. Campbell of the Oak Ridge National Laboratory, which addressed the releases observed during the accident at Three Mile Island Unit 2 (TMI-2) in March of 1979. In that letter, the authors stated, in part, that ". . . the unexoectedly low release of radiciodine in the TMI-2 accident is now understood and can be generalized to other postulated accidents and to other designs of water _ reactors." The
!cw releases observed at Tg11-2 and a call for further. investigation were discussed in a December 21, 1980 letter addressed to President Carter from Governor Babbitt (of Arizona) and other members of the Nuclear Safety Oversight Committee established by the President in the af termath of TMI-2. In the period from 1930 to the present, a truly massive investment of resources has been expended by both government and private industry to improve the knowledge and understanding of the potential release of radionuclides during the postulated accidents at nuclear power plants.
2.2 Summarv of Recent Investigations The American Nuclear Society (ANS) established a Special Committee on Source Terms in June of 1932 to review and evaluate the state of knowledge of ho'v to predict source terms deriving from severe core damage accidents, and to summarize the findings of various investigating organizations. The state of knowledge and associated findings were q a
h 5 Stratton, W. R., Malinauskas, A. P. and Campbell, D. O., letter to NRC Cha;rman,
- 3. Ahearne, August 14, 1980.
i 6
Bab5itt, B., Deutch, J., Goldberger, M., and Lewis, H., Letter to the President from Nuclear Safety Oversight Committee (NSOC), December 1980.
o
.. i 9 -
comoared to those cof WASH-1400. The ANS Committee published its reoort in September 1934 and. summarized the results at a meeting of the Nuclear ' Regulatory Commission in November 1984.
2 The major finding of the ANS Special Committee was, in parttas follows:
" Source items for severe core damage accidents have been overestimated by large factors both in government and industry publications. With a small number of exceptions, estimates of source terms associated with severe core damage accidents can be reduced from estimates in WASH-1400 by more than an order of magnitude to several orders of magnitude. _ The noble gas fission products (kryoton and xenon) are exceptions. Because of their chemically inert character, they do-not undergo the wide range of chemical and. physical' interactions which are the fundamental cause of the reduced release of most fission products. However,' the -
very fact that they are inert also leads to low radiological consequences from their -
release. Soecifically, the Committee finds:
For large dry PWR containments, sufficient information exists to support the calculation of source terms ranging from a small fraction of a percent to no .
more than a few percent of the core inventory of imoortant fission products '
species."
The Committee also reported a number of specific findings relative to other plant tvoes and other findings supporting or qualifying the maior finding. Those findings addressed such areas as containment integrity, thermal hydraulics, fission product transoort and deposition, and important radionuclides and chemical forms.
In recognition of the importance and complexity of the investigation of severe accident source terms, the NRC established an Accident Source Term Program Office (ASTPO) to coordinate the NRC's investigations. A major undertaking of that office _ was the management of an extensive study by Battelle CMumbus Laboratories, tbe results of which were published in a multi-volume report, designated BNil-2104.g In further ~
recognition of the importance and complexity of tr.e subject, ASTPO organized and conducted extensive public peer review meetings during the course of the development of B.\11-2104.
In order to provide an independent peer review by scientists not normally associated with nuc! car power plant studies, ASTPO commissioned an investgation by the American Physical Society (APS) which formed a Study Group on Radionuclide Release from Segere Accidents at Nuclear Power Plants. The APS Study Group completed its reoort in
~
February 1935 and briefed the Nuclear Regulatory Commission in ' larch 1935 on its findings and recommendations.
7
" Report of the Special Committee on Source Terra", American Nuclear Society, September 1934.
8 Gieske, 3. A.. et. al., "Radionuclide Release 11nder Specific LWR Accident Conditions", Bill-2104, Battelle Columbus Laboratories, July 1934.
9 "Draf t Report of the American Physical Society Stu fy Group on Radionuclide Release from Severe Accidents at Nuclear Power Plants", American Physical Society, February 1935.
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The APS Study Group's princioal conclusions were as follows:
"The study group ~ finds that considerable orogress has been made'since publication of the . Reactor Safety Study (NRC 1975) in developing'both a scientific basis and
. calculational ability for predicting the -source term. In' a number of cases, new -
calculations indicate . that the quantity of radionuclides that; would reach the environment is significantly - lower than that calculated in the Reactor Safety Study. This reduction can be attributed to three principal factors: (1) the recognition that reactor containments are stronger than ' assumed in the Reactor Safety Study and therefore fail, -if at all, at later times: (2) inclusion in the modeling of previously neglected physical and chemical phenomena that lead to the retenti.on of fission oroducts; and (3) inclusion of additional sites (suooression pools, ice - beds, auxiliary buildings) that . trap radionuclides more efficiently than previously assumed."
"The study group examhed the chemical and physical phenomena considered by the technical community since the Reactor Safety Study was completed. For most sequences and most radionuclides, these phenomena reduce the source term from that calculated in the Reactor Safety Study." ,
"However, one mechanism that might, Ifor some sequences, increase the radionuclide releases above those calculated in the Reactor Safety Study is the release of non-volatile radionuclides in the core-concrete interaction. It is important to complete the exoeriments now underway to improve our knowledge of the ohysics and chemistry in this crucial area. Moreover, the analyses oerformed in the recent st' dies that we have surveyed have not treated all types of reactors nor all types of containments in equal detail. It is_ impossible to make the sweeoing ,
generalization that the calculated source term for any accident sequence involving any reactor olant would always be a small fraction of the fission product inventory at reactor shutdown. Although further studies may improve this situation, some of the reasons for this inability are enumerated . . .".
The Study Group report went on to discuss details of the above conclusions.
- Subsequent to the publication of the ANS and APS reports, and reports of a number of industry organizations, such as the Industry Degraded Core Rutemaking (IDCOR) program, ; the Electric Power Reseat ch Institute - (EPRD, the Stone and Webster Engineering Corporation, and others, the NRC published a' Draft Report for comment, entitled "Reassessgnt of the Technical Bases for Estimating Source Terms" - NUREG-0956, in July 1985.
The first conclusion provided by NUREG-0956 is:
"The BMI-2104 suite of computer codes reoresents a m @r advance in technology and can be used to reolace the Reactor Safety Study mdpds." g (emohaAs ,
added.)
10 Silberberg, M. et. al., " Reassessment of the Technical Bases for Estimating Source Terms", NUREG-0956 Oraft Report for Comment, U.S. Nuclear Regulatory Commission, July 1985.
1I ibid, p. 3-1.
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TABLE 2-3 PWR SOtJRCE TERMS IN NtJREG-0956 -
(Surry)
Accident V V Sequence TML1Y-Be TMLB'-# TMLIY-e Suhmerged tJnsubmerged AB-y AB-p AB-c S2 0'Y
{rogabiliger 1.7x10-8 6.6x 10-9 3.0x10-7 3x10-6 lx10-6 lx10-10 < 10-10 < 10-10 5.5x10-3 Release 2.5 2.0 12 1.0 1.0 4.5 0.5 Time, lir 24 2.5 Fission Product Fraction of Core Inventory Released to Environment Group Xe-Kr 0.85 1.00 0.S0 I.00 1.00 0.80 0.90 0.15 0.50 I-Br 0.07 0.022 0.0023 0.08 0.40 0.057 0.037 0.000043 0.005 Cs-Rb 0.053 0.013 0.00039 0.03 0.40 0.06 0.087 0.000047 0.0001 Te-Sb 0.055 0.11 0.085 0.025 0.I 2 0.14 0.066 0.00004 0.01 Ba-Sr 0.01 0.053 0.018 0.0022 0.011 0.097 0.076 < 0.00001 0.03 Ru 0.0013 0.0053 < 0.00001 0.00013 0.00065 0.0024 0.0029 < 0.00001 0.00i La 0.0017 0.0002 0.0001 0.00007 0.00035 0.008 0.0075 0.000036 0.0009 A
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WASH-1400 and NUREG-0956 release fraction terms, which apply.to these sequences, is presented in Table 2 4 From this comoarison it can be observed that, excluding noble gases, the source terms using the revised and improved models in- NUREG-0956 are generally substantially lower than those reported in WASH-1400, with the exception of the V-unsubmerged sequence.
NUREG-0956 addresses the analyses for t'vo variations on the V sequence for Surry. the V sequence-includes a postulated unisolated break in the low pressure portion of the low pressure safety injection system (LPIS) outside containment. This sequence constitutes a containment bypass sequence which permits direct release from the reactor coolant system to a contiguous structure. These analyses only apply to Surry and are not applicable to other plants, such as CCNPP.
With regard to the large break loss of coolant (LOCA) sequences B y, AB-4, and AB-e ,
the probabilities of occurrence are so low, i.e., less than ! x 10-10 per reactor year, that they were discarded from further consideration in NUREG-0956. The release fractions for the small break LOCA analysis, S2D, are applicable to other PWRs, as is the case with the TMLB' analyses. The probability of occurrence of these sequences may be different than for Surry, based on plant-specific features. .
3.0 Aonlication of Recent Source Term Knowledge to CCNPP in applying the recent source term knowledge to CCNPP, the aporoach adooted was to utilize the source terms reported in NUREG-0956 for Surry for the "in-containment" accident sequences and develop CCNPP plant-soecific source terms for the containment bypass V sequence.
.The first step in determining the applicability to CCNPP of source terms based on-analyses of,Surry is to compare the plant features. Section 3.1 summarizes a comparison of plant features for Surry, Zion and Calvert Cliffs.
In addition to the comparison of plant features, an engineering evaluation of important accident sequences at Calvert Cliffs was conducted, as summarized in Section 3.2. This evaluation concentrated on sequences which represent major contributions to calculated core me!t frequency. Many of these sequences would result in low source terms owing to the operation of engineered safety features.
Severe accident source terms were included in the analysis of offsite consequences at CCNPP based on a combination of the results of the plant features comparison discussed in Section 3.1 and the ergineering evaluation discussed in Section 3.2. The source terms and their associated probability of occurrence are discussed in Section .3.3.
3.1 Comoarison of Plant Features Table 2-5 lists plant features for Surrv, Zion and Calvert Cliffs which are impoesnt with .
regard to severe accident source terms analysis. The following discussion addresses each feature or group of features.
Features I and "2 Both Surry and Zion are Westinghouse pressurized water reactors (PWRs) whereas Calvert Cliffs is a Combustion Engineering PWR. However, their reactors and reactor coolant systems (RCS) are sufficient!v similiar from a fission product transport standpoint to expect comparable results for the RCS portions of the source term a 1' bwy .
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COMPARISDN OF StfRRV. 710N. AND cat. VERT CLIFFS Pt&NT FEATURES Feature Surry 7 ton Calvert Citffs
- 1. Power (Mwt) 2440 3250 2700
- 3. Containment f ree volume (f t') 1.8 x 10' 2.8 x 10' 2.0 x 10*
- 4. Reinforced or post-tenstoned containmont Retnforced Post-tenstoned Post-tensioned
- 5. Containment operating pressure (psta) 10-11 Atm Atm
- 6. Containment desion pressure (psta) 60 62 65
- 7. Containment fatture pressure (psta) 135 150 140
- 8. In-core Instrumentation penetratton location Bottom Bottom Top
- 9. Cavity coredition (Olden spray f a t ture) Dry Flooded Dry
- 10. Cbncrete type Basaltic Limestone- Sil iceous '"
St. Spray without emergency power No Yes No
- 12. Spray rectrc. Independent of ECCS rectrc. Ves No Yes
- 13. ESF containment unit coolers /f il ters No Yes Yes
- 14. No . o f htDh pressure valves in ECCS discharDe 2*" 4 4 "* ,
- 15. ECCS low pressure Itne break location submerged Yes Not investigated See Note 5
- 16. RHR for shutdown coollog) system inside containment Yes No No
- 17. No. of high pressure valves in RHR/SCS letdown path N/A 2*" 2 '"
- 18. kHR/SCS letdown break location submerged N/A No No IS. ConttDuous structure free volume (ft') 10,000*" 1.4 x 10' 2.5 x 10' Auxtltary Engineered Auntitary' SafeDuards Butiding Building Butiding (1) Two par allel cold legs per loop.
(2) CACO. content comparable to basaltic. ,
(3) Includspu two check valves and MOV in cold leg flow path locked open.
(4) Including one normally open weighted check valve closing on 300 Dpe reverse flow.
(b) HIDn pressure pump suction submerDod. Iow pressure pump discharDe not submerged.
(6) Interlocked to preclude opening with ht0h RCS pressure.
(7) BMI-2tO4 assumed approximately 200.000 ft' (apparently includfog quendh spray pump house and main steam walve house),
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analyses. The reactor power levels for Surry and Calvert Cliffs ar.e similar, allowing for close correlation of analysis features which are strongly influenced by power level.
Features 3 through 7 The containment free volume is an important consideration in aerosol behavior in the containment. The Calvert Cliffs containment free volume is sufficiently similar to that of Surry to permit the application of the results of,anaivses of fission product retention in the Surry containment to Calvert Cliffs. Although Surry is a reinforced containment, whereas Zion and Calvert Cliffs are post-tensioned containments, the most imoortant feature of the containment relative to source terms considerations is the containment failure oressure. All three plants have containment failure cressure well in excess of the loads calculated to be imposed on them in the initial phases of severe core damage accidents. Although the timing of postulated late overpressure failure of the containment would vary from plant to plant, the source terms are expected to be very low for such scenarios at all three plants. The pressures included in the attached tabulation for Surry and Zion are based on NUREG-0956. The Calvert Cliffs data were developed by BG&E.
Feature 8 The Calvert Cliffs plant includes in-core instrumentation penetrations in the top head of the reactor pressure vessel, whereas the Surry and Zion plants have lower head penetrations. However, the source term analysis for Surry and Zion reported in NUREG-0956 did not include the lower vessel head penetrations as sites for early head failure.
Therefore, those analyses are directly comparable to the Calvert Cliffs continuous lower head configuration.
Features 9 and 10 A dry, or nearly dry, lower reactor cavity permits an earlier and more aggressive core / concrete interaction than does a flooded cavity. The overlying water in a flooded cavity also permits the removal of fission products released from the core / concrete interaction. The Surry analyses reported in NUREG-0956 are based on a dry cavity and are thus directly applicable to Calvert Cliffs. Of somewhat less importance is the concrete type. Limestone concrete produces considerably more carbon dioxide during the core / concrete interaction than does basaltic (or siliccous) concrete, leading to a somewhat earlier containment overpressure condition. The NUREG-0956 analysis for Surry was based on basaltic concrete. From a mass and energy release oerspective, it should be reasonably comparable to the siliceous concrete at Calvert Cliffs.
Features 11 through 13 Containment failure time and fission product aerosol removal are greatly affected by the availability of containment sprays. The Surry and Calvert Cliffs systems are compara5te in this regard. It should be emphasized, however, that the severe accident sequences which result in the highest source terms are those in which containment sprays are assumed to be unavailable. Calvert Cliffs is similar to Zion in relation to safety-related containment unit coolers / filters, which are not included in Surry. Analyses for Surry result in higher source terms than would be the case with assumed availability of these systems. Thus, application of the Surry results to Calvert Cliffs, which includes these additional sites for removal of fission product aerosols, is conservative.
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Features II4 through 19 These features are included in
- the attached compilation because they relate to interfacing systems loss of coolant accidents (LOCAs), which historically have been referred to as V sequences since WASH-1400. In the WASH-l'400 and NUREG-0956 analyses of Surry, the V sequence analyzed included the overoressure failure of the low pressure ' portion of the Emergency Core Cooling System (ECCS) located outside containment. Thus, the sequence represented a containment bypass accident. At Surry, the most probable location of the low pressure oloing break would be suSmerged by RCS water and by water from the Refueling Water Storage Tank (RWST). A review of the Calvert Cliffs plant arrangement indicates that a similarly flooded break location is possible. Due to the plant unique valve arrangement at Surry (two check valves and a locked-open motor-operated valve (MOV)) a relativelv high probability was assigned to ECCS system failure. Such a failure mode is much less6 pr pable for Calvert Cliffs, and if it were to occur it would occur in the large (2.5 x 10 ft I auxiliary building, possibly at a flooded break location.
The probability of a failure in the low oressure Shutdown Cooling System (SCS) letdown line at Calvert Cliffs is more likely than the ECCS low pressure injection line due to the ,
presence of only two isolation valves in the SCS line, as opposed to four in the ECCS line. (At Surry, the Shutdown Cooling System is referred to as the RHR system and is' located inside containment, thus a failure in that system does not constitute a containment bypass sequence). Moreover, the ECCS low pressure injection line failure for Surry reoorted in NUREG-0956 is not aoplicable to Calvert Cliffs because of the small contiguous structure 'used in the analysis and the probability of the failure itself (i.e.. number of valves in the line). Thus, a plant-soecific analysis of the interfacing system LOCA sequence is required for Calvert Cliffs because of the substantial difference in plant features from the Surry analysis reported in NUREG-0956.
Based on the above comparison of plant features it is concluded that:
o Calvert ' Cliffs is sufficient!v similar to Surry to allow the aoolication of the results of Surry source term analyses to Calvert Cilffs for accident secuences involving releases of fission products into the containment building (i.e., so called "in-containment" sequences).
o The interfacing system LOCA sequence (V sequence) analysis in NUREG-0956 for Surry is not applicable to Calvert Cliffs, and a olant-specific analysis is required.
3.2 Engineering Evaluation of Accident Secuences at CCNPP 3.2.1 Selection of Accident Secuences S in determining which potential severe accident sequences to fo'cus on, the following factors were considered:
o Contribution to overall core melt frequency as oresented in the Calvert Cliffs Interim Reliability Evaluation Program (IREP) Report (NUREG/CR-3511).
o Sequences which involves a spectrum of the possis!e severe core damage accidents (LOCAs, transients, etc.)
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.1 .
a spectrum of thermal hydraulic and fission oroduct transoort chenomena .
in the reactor coolant system, containment and auxiliary building a ore-existing breach of containment (i.e., containment isolation failure) containment bypass (i.e., interfacing systems LOCA outside containment).
Based on these factors, the following severe accident sequences were selected for evaluation:
o ATWS - Anticioated Transient Without SCRA\1. ATWS sequences contribute 33 percent to the total care melt frequency as reported in the Calvert Cliffs IREP study. All of these sequences result in transient induced LOC As because of either a stuck open relief valve or a rupture of the reactor coolant pressure boundary. Containment sprays are available, thus resulting in low expected source terms for this group of sequences.
o TML - Transients with complete loss of feedwater. T\il sequences contribute 32 percent of the total core melt frequency in the IREP study for Calvert Cliffs. Since the shutoff head of the high pressure ECCS injection pump is less than the relief valve setooint, it is possible that all core makedo may fail if heat cannot be removed from the steam generators. Containment sprays are available, thus resulting in low expected source terms for this group of sequences.
o 59H/S7D" - Small LOCAs with loss of ECCS. These two sequences include ECCS failure in either the injection (D") or recirculation (H) phase. Together these sequences account for 12'6 of the total core melt frequency in the Calvert Cliffs IREP study. In both cases, containment sprays are available.
o 52FH - Small LOCA with failure of containment spray recirculation and high pressure ECCS recirculation. This sequence was a significant contributor to the total core melt frequency in the Calvert Cliffs IREP study (9 oercent) and was also analyzed in the IDCOR program analysis of PWRs. Containment sprays do not function during the time period of postulated fission product release to the containment.
o Station blackout. As analyzed in the Calvert Cliffs IREP study, this sequence is postulated to result in the loss of the power conversion system (PCS) uoan loss of all ac power. Af ter approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, station de power is assumed to be lost due to batterv depletion, resulting in loss of the auxiliary feedwater system (AFW). It reoresents one of the important core melt sequences in IREP (3 percent) and is very similar to the TMLB' PWR accident I
analyzed in WASH-1400, B\il-2104, IDCOR, and other studies. The difference between the traditional T\iLB' and the Calvert Cliffs Station Blackout sequence is the delaved failure fafter about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) of the AFT system.
Phenomenologically, it is the same as the WASH-1400 TMLB'. Containment sprays are not postulated functica for these sequences. ,
o Interfacing system LOCA sequences. As is the case with all PTRs, the Calvert Cliffs olant contains several potential containment bypass flowpaths. The WASH-1400 V sequence was not considered in the Calvert
-9
o.
Cliffs IREP study because of the extremely low assessed probability of the necessary valve failures to expose the low pressure portions of the ECCS to full RCS pressure. The Calvert Cliffs ECCS has 3 check valves and one
.normally closed motor coerated valve (MOV) between the reactor vessel and the low pressure piping on both the low oressure ECCS pump discharge and the high pressure ECCS pump suction in the auxiliary building. Based on the ECCS configurati grounds (1 x 10~g perpn, the V reactor sequence year). was The most dismissed probable in IREP pipe break on orobabilis location in the low pressure oortion of the high pressure ECCS would be submerged by break effluent; for the low pressure ECCS it would likely not be submerged.
A Shutdown Cooling System (SCS) letdown line "V sequence" has been included in this evaluation based upon examination of other containment bypass flowpaths with fewer isolation valves than the ECCS flowpaths and an assessment of potential pipe break locations in these flowpaths which would result in an unsubmerged break condition.
3.2.2 Summarv of Eneineering Eeatuation of Selected Secuences Engineering evaluations were performed for selected sequences to determine which sequences would be bounded by corresconding sequences for Surry, as reoorted in NUREG-09%, and which sequences would require additional or plant-soccific treatment. As discussed below, the CCNPP sequences were found to be bounded by the corresponding Surry sequences in al; cases except the containment bypass sequences.
ATTS In the Calvert Cliffs IREP study, several different Anticioated Transient Without SCRAM (ATWS) sequences were oostulated. They differ with respect to initiating event and subsequent system failures; however, all of them include a major transient fo!! owed by a failure to SCRAM.
Although the combined contribution of ATWS sequences to the total Calvert Cliffs IREP core melt frequency is 33 percent, these sequences do not involve high source terms due to the mitigating effects of engineered safety features.
Baltimore.Cas and Electric (BG&E) has previously committed to imolementing those portions of the ATWS rule apolicable specifically to Combustion Engineering (CE) olants; h, installation of an independent trio system. Additionally, as a result of the Salem ATWS event, BG&E has developed procedures for ATWS events for the Calvert Cliffs Control room ooerators which call for:
- 1) Manual SCRAM.
- 2) De-energization of the motor-generator sets which sucoly control power to the reactor protection system SCRAM circuitry fwhich operates on a de-energize-to-SCRAM concent).
- 3) Manual initiation of emergency boron injection.
BG&E has alto refurbisSed the SCRAM 5rea'<ers at Calvert Cliffs and has orovided improved testing and maintenance procedures. Therefore, the basic SCRAM system failure frequency in the Calvert Clif fs IREP study has been measurably reduced.
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In all postulated AT'VS sequences, the containment heat removal systems, Containment Spray System (CSS) and Containment Air Recirculation and Cooling System (CARCS),
remain available. Although effectiveness of the CARCS may be somewhat reduced 5y the presence of aerosols in the containment atmosphere, the CSS and GARCS system provide redundant and independent means of containment cooling and fission oroduct removal. Two dedicated 100 percent capacity containment spray pumps are installed, along with a nozzle arrangement which provides complete horizontal coverage of the containment cross-section.
Both the CSS and CARCS systems are safety-related with resoect to power suop!!es, controls, etc. and have no unusual system dependencies or coerating characteristics.
Since the CARCS may not be completely effective, only the CSS is assumed to be available during an ATWS accident sequence in this evaluation. From a fission product release viewpoint, ATWS should be a benign event due to continual containment heat removal and fission product scrubbing.
T\1L There are five significant sequences included under this heading, all of which involve a transient-induced loss'of the PCS and subsequent loss of auxiliary feedwater. Since the high pressure ECCS pumos have a relatively low shutoff head of aporoximatelv 1275 osia, it is conservatively assumed that it may not be possible to guarantee cooling of the core by " feed and bleed" with the primary relief valves cycling followir.g a loss of secondary heat removal. Therefore, it is postulated that ECCS makeup will fail. In fact, ooerator actions are possible that would force the relief valves to remain open, possibiv allowing the reactor pressure to decrease below the shutoff head of the high oressure ECCS pumps. However, no credit was given for them in the IREP study. For all of the Ti1L cases, CSS is assumed to be effective in substantially reducing source terms.
S H/S,D" la both of these sequences, a small LOCA (S is postulated to occur, including reactor shutdown via SCRA\1 and initiation of AFW.2) In the 2 case of 5 H, high oressure injection is successful as well, but in the case of S 0", it is not. Therefore, in the case of 5 7 D", loss of reactor water inventory is assumed2to be immediate. In the case of S.,H, the ECCS fails when the RWST depletes and the switchover to recirculation occurs. This switchover is expected to occur between 4 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter the start of the LOCA depending on the size of the break. In either case, due to the lack of makeuo to the reactor vessel, core uncovery and fuel damage are postulated to occur. Because the CSS is assumed to be effective, both of these cases are relatively benign from the standpoint of fission product releases.
S,FH In this sequence a small LOCA (52) is postulated to occur which is followed by successful reactor shutdown via SCRAtt and initiation of AFW and high pressure ECCS. This 4 provides for decav heat removal via the steam generators and maintenance of reactor l vessel water inventory. When the RWST depletes and switchover to recirculation occurs, the containment scrav recirculation mode and high pressure ECCS recirculation mode are assumed to fail (F and H resoectively). This switchover is expected to occur between 4 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter the start of the LOCA depending on the size of the break. Due to the lack of makeup to the reactor vessel, core uncovery and fuci damage are costulated to occur.
-w Since the CSS is assumed to be inoperable due to failure of the recirculation oortion of the system, the S.,FH sequence is not considered benign with respect to fission product releases. The CARCS is available to remove heat and fission oroducts from the containment atmosphere, but because the effectiveness of the CARC3 may be somewhat reduced by aerosol loading, no credit is taken for its operation.
Station Blackout in this sequence, a loss of offsite oower initiating event is postulated to' occur, fo!! owed by a SCRAM and subsequent loss of onsite emergency power. The oower conversion system would not be available without offsite ac power. The AFT system continues to provide makeup to the steam generators and decay heat is removed from the RCS via the steam generator relief valves. At some time subsequent to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the sequence, station de power is postulated to be lost due to battery depletion and the AFW system fails on loss of de control power. Core boiloff, uncovery, and subsequent fuel damage are postulated to occur assuming that neither onsite nor offsite ac power is restored.
Because ac nower is assumed unavailable, no containment cooling systems (CSS, CARCS) are operable during this sequence. This accident is very similar to the definition of TMt.B' found in WASH-1400 and other severe accident risk and source term studies. The -
difference in this case is the delayed failure of AFW by at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, which takes credit for the availability of de power for a given period following loss of all ac power.
Without ac power, and assuming eventual deoletion of de power, a core degradation and melt progression are postulated to occur. However, several important factors affecting accident probability and timing deserve consideration. The assumption that the station batteries will orovide de power for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is derived from the fact that dc loads under station blackout conditions are substantially smaller than under design basis LOCA conditions. The design basis capacity for the de oower suoply is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. In addition, a fully qualified reserve battery with an additional 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> capacity is also available under station blackout conditions. The IREP study assumed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and actual test discharges for the Calvert Cliffs battery system have indicated that the battery will operate successfully for several hours following loss of all ac oower. The length of time that dc power is available is important because it offers more time to restore offsite power and/or emergency diesel power. Additionally, the Calvert Cliffs olant has a separate NRC-approved source of offsite power which was connected during plant construction and is still used to provide hotel services to administrative areas. This power originates with the Southern Maryland Electric Co-Ooerative (SMECO). Although not safetv-related in terms of distribution or control, it is available and is in addition to the normal offsite power supply. Use of the SMECO power supply could provide a means of recharging the station batteries. Extension of de power operation would provide extra time to restore ac power and would extend the time of core uncovery if ac power could not be restored. Taking credit for the SMECO power sucoly would reduce the calculated probability of station blackout for Calvert Cliffs below that used in IREP.
V (Interfacing System LOCA Bvnassine Containment) , 1 1
In this sequence, a ploe or comoonent directly connected to the reactor coolant pressure boundary (or oart of that boundary) is assumed to break outside containment. Because of the isolation capability provided for all oloing systems penetrating containment, these sequences require multiple valve failures to occur. System designs vary from plant to plant, necessitating attention to specific plint features which might result in an interfacing systems LOCA byoassing containment.
The following table presents the results of an engineering survey carried out by Stone and Webster Engineering Corporation and the Calvert Cliffs plant staff. Several potential V l
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sequence flowpaths were identified and evaluated, taking into consideration piping size and configuration, the location of the pioing code class changes (high pressure to low pressure), the number of valve failures required to reach the code class change, the room elevation of the possible break, and other pertinent features. In addition, the ootential for submerged break locatiens within flood protected compartments was evaluated. -
System / Pipe No. of Valve Elevation of Flowoath Size Failures Break Within Room HPSI 6 in. 4 (Notes 1&2) Low LPSI 6 in. 4 (Note 2) High SCS Letdown 14in. 2 High CVCS Letdown 2 in. (Note 3) High CVCS Charging 2 in. 3 (Note 4) Low Note 1: HPSI pumo discharge oiping is code service class CC (L85 psig ratingi or class DC (1600 psig rating).
Note 2: Includes one check valve 'vhich is normally open but which will close if .
reverse flow exceeds 300 gpm.
Note 3: CVCS letdown flow must pass through 2 air-operated valves.
Note 4: CVCS charging pump discharge lines are code service class CC (2435 osig rating). The 3 valves are all check valves.
Based upon the results of the above survey, the interfacing system LCCA considered most !!miting is a break in the SCS letdown line resulting from the failure of 2 high pressure containment isolation valves. This scenario also provides the largest possible leakage path (h,14 in diameter pipe) and an unsubmerged break location.
It should be noted, however, that the two valves in the SCS letdown flowpath which must fail in order to expose the low pressure portion of the system to the RCS pressure are both f!cx-wedge MOVs. - A survey of nuclear power plant operating histories and discussions with several large valve manufacturers have not revealed a single incident of large Icakages past the disk of this type of valve, let alone catastrophic disk failure. For the purposes of this study, however, a conservative disk failure rate characte-istic of check valves has been used to assign a probability to this event.
The potential also exists for a failure in the low pressure portion of the SCS at a oumo seal, which would result in a submerged break in the flood protected pump compartment. Therefore, this sequence was also chosen for analysis to provide a range of V-sequence outcomes. The pump seat leak was designated as the low range V sequence, and the 14" line break was designated as the high range V sequence based on the relative sizes of the expected source terms.
- Because the flowoaths involve LOCAs in the auxiliary building, the containment and its heat removal transport and systems retention are directiv in the bypassed.
auxiliary building.Attention is thus This building is focused large (2.5onx fission org) duct 10 ft and consists of several elevations and a large number of compartments. It also contains a seismically supported fire protection sprinkler system. The analysis of the source terms for these sequences is summarized in Section 3.4.
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3.3. Summarv of Results-As a result of the above evaluations, the fo!!owing sequences were included in the CCNPP source term analyses:
Sequence Descriotion Origin of Source Term Used -
TMLB'-6 e Station blackout with early NUREG-0956 overpressure break of the containment TMLB'-4 Station blackout with failure
- NUREG-0956 to isolate the containment S2 D-? Small break LOCA with NUREG-0956 containment break due to H2burn V Interfacing system LOCA CCNPP bypassing containment Plant-Soecific The probability of occurrence of the severe accident sequences included in the CCNPP offsite dose analyses is summarized in Table 2-6. As noted in this table, the CCNP,P probabilities were derived from a combination of data from the IREP study for CCNPP and containment event tree probabilities reported in NUREG-0956. The V sequence probability is based on the orobability of an RHR - V sequence as reported in the Zion PRA. Table 2-7 lists the severe accident source terms utilized in the analysis of offsite doses at CONPP.
.The first three listed "in-containment" sequence source terms were taken from the NUREG-0956 analysis for Surry, based on the determination that these data are directiv applicable to CCNPP. The two V sequence source terms are Calvert Cliffs plant specific. The range of the V sequences possible at CCNPP was represented by " low ,
range" and "high range" source terms. The low range source term comprises the complete release of the noble gas core inventory (xenon and kryptoni, with negligible release of the volatile and non-volatile fission product groups. This source term is judged to be applicable to the low-range V sequence, involving a SCS pump seal rupture resulting in a small size leak into a flood-protected oumo comoartment (i.e., submerged break).
The high range source term was calculated using a detailed analysis of a 14 in, diameter SCS pipe break (unsubmerged).
3.4 Descriotion of Plant-Soecific Contain nent Svoass Sequence Analysis The auxiliary building at CCNPP has an extensive fire suporession system, including sprinklers on all of the levels in the potential accident discharge flowpaths to the environment. These sorinklers are individually set to discharge water when a local temperature of 212 F is reached. There are soproximately 300 sorinklers on Elevation 3 5', which is where the 14" SCS oice break is costulated to occur. The activation of these sprinklers markedly affects the subsequent thermal hydraulics In the auxiliary building.
Initially, 63 of the sprinklers are calculated to actuate based on the temocrature transient associated with the blowdown. During the course of the accident, a total of 140 sprinklers were calculated to be actuated. A minimum water supply of 600,000 gallons is avaliable to the fire suppression system. Based on a design flow rate of 20 gpm per sprinkler, the water supply was calculated to be available until after the complete release of fission products at the break location.
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TABLE 2-6 ^^
PROBABILITY OF OCCIJRRENCE OF SEVERE ACCIDENT SEOtJENCES AT CCNPP (Per Reactor Year)
Mode of CCNPP Release from Probability. Cont. Event Probability Sequence Containment of Event Tree Probability of Release TMLB'-5 e Early 1.5 x 10-5(a) 5 x 10-3(c) 7.5 x 10-8 Overpressure TMLB'-B Isolation 1.5 x 10-5(a) 5 x 10-3fdI 7.5 x 10-8 Failure 52 D-y Early 1.0 x 10-4(b) ! x 10-3(c) 1.0' x 10 ~7 Overpressure V Bypass Via 1 x 10-7 N/A 1 x 10~7 Interfacing Syst. LOCA (a) Based on combining the 4.4 x 10-6 probability of a TMB' sequence with the 1.1 x 10-5 probability of an S Fil sequence from the-IREP Study for CCNPP. 2 (b) Based en combining the probabilities of all core melt sequences with containment sprays operating (ATWS, transients with loss of FW and S2 0"I811) 2 i r the IREP Study for CCNPP.
(c) Based on NUREG-0956.
(d) Based on EPRI/NSAC analysis for Oconee (NSAC-60 Oconee PR A, June 1984).
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TABI.E 2-7 SEVERE ACCIDENT SOURCE TERMS AND PROBABILITY OF OCCURRENCE FOR CCNPP Range of V Sequences Accident Secuence TM LB'-a cI ^ TMLB'-B(a) s20-7 f} l"*(b) liigh(c)
Probability, Per 7.5 x 10-8 7.5 x 10-3 1 x 10-7 I x 10-7(d) 1 x 10-7(d)
Reactor Year (See TABLE 2-6)
Time of Re!, hr. 2.5 2 2.5 > 10f *I 4 Duration of Rei, hr. 10 10 3 1 3 Fission Product Grotn> Fraction of Core Inventory Released to Environment Xe-Kr 0.85 1.00 0.50 1.00 1.00 I-Br 0.07 0.022 0.005 <0.001 0.039 Cs-Rb 0.053 0.013 0.0001 <0.001 0.033 Te-Sb 0.055 0.11 0.01 <0.001 0.042 Ba-Sr 0.01 0.053 0.03 <0.0001 0.0009 Ru 0.0013 0.053 0.001 <0.0001 <0.000i La 0.0017 0.0002 0.0009 <0.0001 < 0.0001 (a) Release fractions from NUREG-0956.
(b) Low range of potential V sequences inchxles small leakage paths (e.g., 0.1 ft 2) at a pump seal submerged in a flood-orotecte<1 pump compartment.
(c) liigh rapge of potential V sequences is based on a 14 in. diameter SCS pipe opening - unsubmerged.
(d) 1 x 10- is the total probability of the r would be sihstantially lower than ! x 10. yge of V sequences for SCS. The probability of experiencing the high range source (e) . Time of release is a function of postulate <l size of opening in SCS.
(f) Analyzed in CRAC2 with a 2 hr. release duration. -
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e ,
The fission product aerosol spray removal resulting from the operation of the sprinklers.
was included in the analysis of the source term for the postulated'unsubmerged 14" SCS pipe break. The anaivsis of aerosol spray removalgequired some adjunct calculations in conjunction with the NAUA - 4 computer program.
The method of analysis of the unsubmerged SCS ploe break V sequence is summarized below:
The MAAP-2.0B computer program 15 5vas utilized to analyze the accident progression within the RCS and the containment building. The results of this analysis provided the event timing and the mass and energy releases at the postulated SCS ploe break location.
The thermal hydraulics portion of the analysis of the accident proggssion in the CCNPP auxiliary building utilized the THREED-ST computer program. THREED-ST was previously utilized in 'the performance of extensive thermal hydraulic analysis of the Surry plant, as reported in Chapter 6 and Apoendix B of the report of the ANS Special Committee on Source Terms. Additional engineering calculations were performed, as adjuncts to the THREED-ST analyses, in order to address such issues as determination of the effects of the size distribution of water droplets from the fire suppression sorinkler ,
system.
Based on the results of the THREED-ST analvsis of the high energy line break in the SCS in the auxiliary building, it was assumed that the roof vents would ooen within a few seconds of accident initiation due to the rapid orejsure rise. The 2 areas of the resultant roof-level opegings to the environment are 30 ft and 177 ft . In addition to the roof vents, a 50 ft opening was assumed to occur, at aoproximately the same time, in the metal truck door leading into the refueling bay. These openings not only provide leakage pathways to the environment, but they also affect the thermal hydraulics aspects of the accident sequence by creating circulation paths within the auxiliary building.
A 14 The Stone and Webster Engineering Corporation's version of the NAUA-4 comouter program was utilized in the analysis of aerosol behavior and release from the ,
auxiliary building. The analysis is similar to the Surry analyses reported in the ANS report.
15
.. Modular Accident Analysis Program - MAAP-2.0B", industry Degraded Core Rulemakin.g (IDCOR) Program, (In Publication by Atomic Industrial Forum.)
16 "A Subcompartment Transient Resnonse Code (THREED-ST)", Stone and Webster Engineering Corporatinn Computer Program NU-136 (Unpublished).
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The distribution of volatile fission product species in the two CCNPP V secuence source terms is as follows:
V-High Range Source Term Fraction of Core inventory RCS Soecies & SCS Auxiliarv '3uildine Environment Csl 0.167 0.794 0.039 CsOH 0.206 0.756 0.033 Te 0.116 0.342 0.042 V-Low Range Source Term Fraction of Core Inventerv ,
RCS Water in Soecies & SCS Pumo Como. Auxiliary Building Environment Csl 0.167 0.315 0.017 less than 0.001 CsOH 0.206 0.777 0.016 less than 0.001 Te 0.116 0.365 0.013 less than 0.001 The release of non-volatile fission product soecies in the V sequence is exoected to be extremely low. Reactor vessel meltthrough would occur with the RCS deoressurized, resulting in a low pressure release into the containment (which would be at acoroximatelv atmospheric pressure). The flowpath for the fission products released from the corc/ concrete interaction would then he bac'< through the opening in the lower vessel head, through the vessel and upper vessel Internals, through the RCS piping to the SCS pioing, allowing substantial opportunity for retention en route to the break location 'vith subsequent additional retention in the auxiliary building.
The in-containment sequence source terms, obtained from NUREG-0956, and the CCNPP served as inputs to the plant-specific analysis of offsite V sequence doses source utilizing the C' TACterms, as reported 2 comouter program in w Table y, ith CCNPP site meteorological data. The results of these analyses are presented in Attachment 1.
4.0 Summarv in summary, the new material circumstances, as considered in this application for exemotion, include:
I (1) replacement of the Reactor Safety Study (TASH-1400) results and methods of Ritchie, L. T., et. al., "CRAC2: Calculations of Reactor Accident Consequences, Version 2', NUREG/CR-2336, Sandia National Laboratories, February 1933.
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a analysis.with results and methods of analyses representing a major advance in the technology of analysis of severe accident source terms, as reported in NUREG-0956; '
(2) development of CCNPP plant-soecific source terms for the containment Sypass
~(interfacing sy' stems LOCM " Event V" sequences;
.(3) . use of CCNPP site-soecific meteorology for offsite dose calculations; (4) recognition of the inherent strength of the containment building; and (5) recognition of the inherent capability of plant systems and structures for the retention and removal of fission product soecies.
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