ML19318A145
| ML19318A145 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/13/1980 |
| From: | Wetterhahn M CONNER, MOORE & CORBER, Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18082A579 | List: |
| References | |
| NUDOCS 8006180513 | |
| Download: ML19318A145 (79) | |
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UNITED STATES OF AMERICA s '. NUCLEAR REGULATORY COMMISSION M fore the Atomic Safety and Lice..sina Board i In the Matter ol )
i
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PUBLIC SERVICE ELECTRIC AND ) Docket No. 50-272 GAS COMPANY, et al. ) (Proposed Issuance
) of Amendment to Facility (Salem Nuclear Generating ) Operating License ,
Station, Unit 1) ) No. DPR-70) l l
LICENSEE'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM OF AN INITIAL DECISION l Public Service Electric and Gas Company, et al.,
Licensee in the captioned proceeding, in accordance with 10 C.F.R. 52.754 and the Atomic Safety and Licensing Board's Order of May 9, 1980, hereby submits the attached proposed findings of fact and conclusions of law in the form of an initial decision.
Respectfully submitted, l l
CONNER & MOORE g 'o Mark J. Wetterhahn '
Counsel for the Licensee Of Counsel: j Richard Fryling, Jr., Esq. l Assistant General Solicitor l Public Service Electric and l
-Gas Company l 80 Park Plaza, T5E l Newark, New Jersey. 07101 June 13, 1980 t
800618060
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UNITED STATES OF AMERICA NUCLEAR REGULATORT COMMISSION
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ATOMIC SAFETY AND LICENSING BOARD 1
l Gary L. Milhollin, Esq., Chairman Mr. Frederick J. Shon, Member Dr. James C. Lamb, III, Member In the Matter o'f )
)
PUBLIC SERVICE ELECTRIC AND ) Docket No. 50-272 GAS COMPANY, et al. ) (Proposed Ist,uance of
) Amendment to Facility (Salem Nuclear Generating ) Operating License Station, Unit'l)- ') No. DPR-70)
July , 1980 APPEARANCES MARK J. WETTERHAHN, Esq., of Conner & Moore, Washington, D. C., and s RICHARD FRYLING, JR., Esq., of Public Service Electric and Gas Company, Newark, New Jersey, for the Public Service Electric and Gas Company, et al.
MENASHA J. TAUSNER,, Esq. and SANDRA T. AYRES, Esq., Assistant Deputy Public Advocates, State of New Jersey, for Mr. and Mrs. Alfred C. Coleman, Jr. ,
l CARL VALORE, JR., Esq., of Valore, McAllister, Aron and Westmoreland, Northfield, New Jersey, for the Township of Lower Alloways Creek.
REBECCA FIELDS, Esq., Deputy Attorney General, Department of Law and Publi.c Safety, for the State of New Jersey. l l
JUNE D. MACARTOR, Esq., Deputy Attorney General for the State of Delaware.
JANICE E. MOORE, Esq. and WILLIAM D. PATON, Esq., Office of the Executive Legal Director, U. S. Nuclear Regulatory Commission, Washington, D. C. for the NRC Staff.
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q INITIAL DECISION (AMENDMENT TO OPERATING LICENSE)
PRELIMINARY STATEMENT
- 1. This initial decision involves the application filed on November 18, 1977, with the Nuclear Regulatory Commission ("NRC") by Public Service Electric and Gas Company ("PSE5G" or " Licensee") for itself and' as agent for !
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, the other owners, Atlaltic City Electric Company, Delmarva Power and' Light. Company, and Philadelphia' Electric' Company, for amendment of Facility Operating License -No. DPR-70 for Salem Nuclear Generating Station, Unit No. 1 (" Salem Unit 1" or " facility") located in Salem County, New Jersey. The amendment would revise the provisions of the Technical Specifications, Appendix A to Facility Operating License DP R-70 , to permit the substitution of new spent fuel storage racks to ' increase the fuel storage capacity from 264 to 1170
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fuel assemblies in the spent fuel pool of the facility.
- 2. On T Sruary 8, 1978, the NRC published in the Federal Register (43 Fed. Reg. 5443) a notice of " Proposed l Issuance of Amendment to Tacility Operating License" con-cerning the proposed change. In response, thereto, three petitions for a hearing were submitted. This Atomic Safety and Licensing Board (" Licensing Board" or " Board") was con-stituted to rule on the petitions and later to preside over 1/
the proceedings.-- The Board in this proceeding has been J/ See Establishment of Atomic Safety and Licensing Board to Rule on Petitions dated March 16, 1978 and Notice of Hearing on Amendment of racility Operating License dated April 26, 1978.
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reconstituted on two occasions.--2/ After a prehearing con-ference held on May 18, 1978, the Atomic Safety and Licensing Boar (. admitted two intervenors, Lower Alloways Creek Township
(" LACT") and Mr. and Mrs. Alfred C. Coleman, Jr. ("Colemans")
as parties.- Requests to participate pursuant to 10 C.F.R.
S2. 715 (c) were received from the States of New Jersey and 4/
Delaware- and were granted by the Board. All or part of
^ Contentions 2,'6, 9 and 13 of the Colemans and 1 and 3 of LACT were admitted as issues in the proceeding.
- 3. On January 19, 1979, the NRC Staff transmitted its Safety Evaluation Report ("SER") and Environmental Impact
- i Appraisal ("EIA") to the Board and parties. Pursuant to the Board's Order Following Special Prehearing Conference dated ;
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May 24, 1978, discovery in this proceeding which had begun i after the prehearing conference ended on February 9, 1978, three weeks after publication of the SER and EIA.
- 4. On February 27, 1979, the Licensee. moved the Board for summary disposition of the contentions filed by LACT and 4
the Colemans. The Board' granted the motion as to LACT 5/
Contention 3 and the Colemans ' Contentions 9 and 13..-
_ 2/ Notices of Reconstitution of Board dated March 8, 1979 and June 27, 1979.
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3/ A petition filed by the Sun People ' Alternate Energy Advocates was denied by the Order Following Special Prehearing Conference dated May 24, 1978 at 2-3.
4f See Memorandum and Order dated. April 26, 1978 at 15.and Order Following Special Prehearing Conference dated May 24, 1978 at 2.
_5/- Order dated April 30, 1979. One member of the Board would have granted summary disposition on all issues
- in the proceeding.
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- 5. The record of this proceeding consists of the transcripts of the prehearing conferences held on bhy 18, ;
1978 (Tr. 1-120), March 15, 1979 (Tr. 121-316), March 16 (Tr. 317-365), and the evidentiary hearings held on May 2-4, 6/~ ;
1979 (Tr. 317-918), July 10-11, 1979 (Tr. 919-1351), and April 28-30, 1980 (Tr. 1352-1808), the transcript of an in camera session of the proceeding held on May 3, 1979 to
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7/. l discuss information proprietary to one of Licensee's vendors i and the exhibits which were received in evidence listed in 8/
Appendix A hereto. Numarous limited appearance statements were received. In response to a Board request, the Staff ;
responded directly to a number of the questions raised and 9/ l statements made; the Board is satisfied that the Staff has '
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_6_/ Inasmuch as the reporter has not unambiguously numbered the transcript pages 317-365, any subsequent reference to those pages should be understood to refer to the evi-
- dentiary hearing rather than the prehearing conference.
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-- 7/ The transcript of the in camera session has been received 1
by the Licensee and Exxon Nuclear Company and only three pages were ultimately determined to involve proprietary l matters (Tr. 704-05),.. The Board has determined that, along with Exhibit 3, the designated pages, g camera Tr. 22, 36 and 59, should be withheld from public dis-closure.
t 8/ As a result of Board questions and comments, certain information was submitted by the NRC Staff and Licensee I regarding occupational doses. No objection to the re- l ceipt of this material has been made by any party or participant. As described in detail in Appendix A, this material is received in evidence. Moreover, by order )
dated March 13, 1980, the Board took official notice of ;
a document entitled " Fact Sheet, the President's Program i on Radioactive Waste Management" issued on February 12, i 1980. !
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_9/ See, for example, the July 26, 1979, letter from Staff counsel Janice Moore to Mr. Marvin Lewis.
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responded appropriately. In addition, after the occurrence of the Three Mile Island accident, the Board asked the parties for a presentation regarding the effect on the Salem Unit 1 spent fuel pool if such an event were hypothesized to occur at. Salem Unit 1 and the relative consequences of a
" gross loss of water" scenario. These matters are addressed, infra.
'6. Any. proposed findings of fact or conclusions of law submitted by the parties, hereto, which are not incorporated directly or inferentially into this initial decision, are
- herewith rejected as being unnecessary to the rendering of this initial decision.
FINDINGS OF FACT REGARDING MATTERS IN CONTROVERSY INTRODUCTION Scope of the Board's Review
- 7. This proceeding involves an amendment, which is limited in scope, to an existing full power,, full term operating license ' granted by the NRC in 1976. This Board has recognized the limited nature cf this prcceeding and has comported itself accordingly. The Board believes it is j necessary to give some exposition as to its views regarding !
1 its role as defined by the Commis'sion. This Board has I recognized through this proceeding that it has not been 1
established to conduct a re-review of the decisions reached I during the Commission's operating license review, from either a safety or an environmental standpoint, or, as a l
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corollary, to question the Commission's decision to issue an operating license for the facility.
- 8. This position is in accord with that taken by other Boards in similar situations which have held that they are not required to review the design of all plant systems or reopen environmental questions. considered at the construction and operating license stage, , such as the need for power.
See-Portland General Electric Company (Troj an. Nuclear Plant) ,
ALAB-531, 9 NRC 263, 266, n.6 (1979) and Northern States Power Company (Prairie Island Nuclear Generating Plant, Units 1 and 2) , ALAB-455, 7 NRC 41, 46, n. 4 (1978), remanded on other grounds, sub nom. New England Coalition on Nuclear 10/
Pollution v. NRC, 602 F.2d 412 (D.C. Cir. 19 7 9 ) . '-- We also 1
did not consider it necessary to re-review all aspects of )
the design of the spent fuel pool and associated structures and systems. The Board sees its review as limited solely to any effect that the proposed action, i.e., the installation and use of the new racks, would have on th'ese systems.
Of course, this Board is further limited to deciding the contested 11/
issues before it.
g Cf,. Dairyland Power Cooperative (La Crosse Boiling Water Reactor), LBP-80-2, 11 NRC 44 (1980).
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See Detroit Edison Company (Enrico Fermi Atomic Power Plant, Unit 2), LBP-78-10, 7 NRC 381, 386 (1978). The l Board is, however, mindful of 10 C.F.R. 52.760a and has posed a number of questions, including some related
' to TMI-2.
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. 9. The Board has consistently applied these standards during this proceeding. For example, it rejected reconsidera-tion of accidents or other matters whose analyses would be
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unchanged by approval of the requested action. - Furthermore, in reviewing cert'in a aspe ts of this proposed change, the
, . incremental impact was examined, rather than attempting to establish an absolute basis for licensing as was done by the
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- commission at the operating license stage. This is in -
accord with the directive of the Appeal Board in Prairie Island that a board need only consider whether the amendment itself would bring about significant environmental conse-quences beyond those previously assessed. 7 NRC at 46, n.4.
The Board's Duties Under the National Environmental Policy Act
- 10. An issue that must be dealt with preliminarily is i whether alternatives to the proposed action need be considered I j
at all. The National Environmental Policy Act of 1969, l S102 (2) (c) , 42 U.S.C. 54332 (2) (c) (NEPA) provides in pertinent part that "all agencies of the Federal, Government shall - l
. . . (C) include in every recommendation or report on pro-
- posals for legislation and other major Federal actions sig-i nificantly affecting the quality of the human environment, a detailed statement by the responsible official on - (i)
( the environmental impact of the proposed action."
12/ Order Following Special Prehearing Conference dated May 24, 1978 at 5-6.
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- 11. The Staff performed an environmental evaluation of the proposed modification pursuant to NEPA and issued the EIA on January 15, 1979 (Exhibit 6C) . The EIA describes and i 1
evaluates the Salem f acility, its need for increased spent 1
fuel storage capacity, environmental impacts of the proposed modification, environmental impact of postulated accidents, alternatives.Lfor spent fuel storage, and cost-benefit balance of the proposed modification. Under the heading " Basis and Conclusion for Not Preparing an Environmental Impact State-ment," the EIA states:
We have reviewed this proposed facility modification relative to the require-ments set forth in 10 CFR Part 51 and the Council of Environmental Quality's Guidelines, 40 CFR 1500.6 and have ap-plied, weighed, and balanced the five factors specified by the Nuclear Regulatory Commission in 40 CFR [ Fed.
Reg.?} 42801. We have determined that the proposed license amendment will not significantly affect the quality of the human environment and that there vill be no significant environmental impact attribut,able to the proposed action other than that which has already been predicted and described in the Commis-sion's Final En'vironmental Statement for the facility dated April 1973.
Therefore, the Commission has found .
that an environmental impact state-ment need not be prepared, and that pursuant to 10 CFR 51.5 (c) , the issuance of a negative declaration to this effect is appropriate (Exhibit 6C at 27).
This conclusion is fully supported by the EIA and by the re-mainder of the evidence of record (Exhibit 6C, particularly.
at 5-12 and Exhibit 1C at 6-12) .
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. 12. Initially, this Board does not believe that this is a " major" action. It is not one of those actions which the Commission has determined to be, prima facie, major and which therefore require the preparation of an envirrenmental l 13/
impact statement. Even were we to assume that this action ,
i could be considered a " major" one, the Board agrees with the .
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. Staff that the evidence of record establishes that the in-cremental' impacts associated with the construction, installa-tion of the racks and operation of the spent fuel pool with l the new racks will not significantly affect the quality of the human environment, i.e., will not significantly increase the environmental impact of the Salem Unit 1 facility. This finding is consistent with the decisions of other licensing boards which have addressed this matter. The only action being taken is for the storage of additional spent fuel at the Salem Unit 1 facility. As part of the operating license, permission has already been granted for the storage of one-and-one-third cores resulting from the operation of the unit during its lifetime. Approval of the amendment will not result in the generation of any additional spent fuel.
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- 13. The action by the Commicsion in adopting a State-ment of Interim Policy on Nuclear Power Plant Accident 13/ 10 C.F.R. 551. 5 (a) .
14/ There is also no indication that there is any conflict in the utilization of a scarce resource such as to trigger the environmental review process. See Dairy- ,
land Power Cooperative, n.10, supra. l l
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Cohsiderations under NEPA (" Statement of Interim Policy")-~
subsequent to the close of the record in this proceeding does not by itself or considered in conjunction with the
, other impacts associated with the construction, installation
'auud use of the new racks require the preparation of an 16/
environmental 4.mpact statement in this proceeding.--
Essentially, the new Commission policy announces the with-drawal of the proposed' Annex to. Appendix D of 10 C.F.R. Part 50 and requires consideration of core melt accidents in ongoing and future construction permit and operating license proceedings in an applicant's environmental report and the Staff's final environmental statement where a final environ-mental impact statement has not yet been published. The
, Statement of Interim Policy does not require the preparation
- s. of a new or supplemented impact statement where a final state-
- ment already has been published. Neither does the new l I
policy require the preparation of an EIS where none had previously been re* quired.- In other words, there is nothing i in the new policy which s'uggests that the Commission intended )
to transform an otherwise. environmentally nonsubstantive amendment, i.e., one not requiring the preparation and '
15/ The Statement was approved by the Commission for publication in the Federal "?gister at its meeting on May 15, 1980 and published in the Federal Register on June 13, 1980 (45 Fed. Reg. 40101) .
16/ Nor does it require further consideration of " Class 9" accidents as that term was defined and utilized by the Board prior to the issuance of the Statement of Interim Policy. See n.34, infra.
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publication of an environmental impact statement under the criteria' contained in 10 C.F.R. Part 51, (such as the licensing of an expanded spent fuel pool) into a " major Federal action ;
significantly affecting the quality of the human environ-ment." To interpret the Statement of Interim Policy dif-4 ferently would have 'tdun effect of requiring the preparation of an. environmental impact statement.for every minor amend-ment to an operating. license, a " bootstrapping" result clearly not intended. l
- 14. In the Statement of Interim Policy, the Commission has expressly stated that Class 9 issues shall not be inter-jected into ongoing proceedings, absent special circumstances that indicate a greater risk to public health and safety than previously anticipated:
. It is expected that these revised treat-ments will lead to conclusions regarding the environmental risks of accidents similar to those that would be reached by a continuation of current practices, particularly for cases involving special circumstances where Class 9 riski have been considered by the staff, as described above. Thus, this change in policy is not to be construed as any lack of con-fidence in conclusions regarding the en-
, vironmental risks of accidents expressed in any previously issued Statements, nor, absent a showing of similar special cir-cumstances, as a basis for opening, re-opening, or expanding any previous or ongoing proceeding. 1]/
12/ 45 Fed. Reg. at 40103 (emphasis supplied) .
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As a factual matter, there is nothing in the record of this proceeding which establishes such "special circumstances."
Accordingly, there is no need for this Board to consider the environmental consequences of what were formerly referred j 1
~ to'as " Class 9" acciden'ts, nor is the Staff required to amen'd or supplement its FES issued at the operatiing license stage.or to issue an environmental impact sta.tement in
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~ conjunction with its consideration of the instant licen' amendment solely to discuss " Class 9" accidents. We there-fore affirm the Staff's determination to make a negative declaration pursuant to the Commission's regulations, 10 C.F.R. SS51. 5 (c) (1) and 51.7, and pursuant to the Council on 18/
Environmental Quality Guidelines, 40 C.F.R. 1500.6(e).
- 15. The Commission's rules do not require that an environmental appraisal accompanying a negative declaration consider alternatives.--19/ Thus, it is not necessary to 18/ 10 C.F.R. 51.'5 (c) (1) provides in pertinent part:
[I] f it is detdrmined that an environ-mental impact statement need not be pre-
. pared . . ., a negative declaration and environmental impact appraisal will, . . .
be prepared . . . .
'19/ 10 C.F.R. 51.7 provides in pertinent part:
(a) Negative declarations. The negative declaration required by 551.5(c) will be prepared prior to the taking of the as-sociated action and will state that the Commission has decided not to prepare an environmer,tel impact statement for the particular action and that an environ-mental impact appraisal setting forth the basis for that determination is available for public inspection. Nega-l (Ft. 19/ cont. on next page) 5 y .-4 , , - , . , - - , , . -..,_v
discuss or choose among available environmental alternatives cn the basis of the record in the proceeding. --20/ In our findings , supra, we have determined that the adverse environ-mental. impacts associated with this license amendment will 9/, (continuad) tiva declarations will be published and made- publicly available in accordance with 5551.50 (d) and 51.55. Lists of negative declarations will be maintained and made publicly available in accordance with 551.54 (b) .
(b) Environmental impact appraisals. An environmental impact appraisal will be prepared in support of all negative decla-rations. The appraisal will include:
(1) A description of the proposed ac' ion; (2) A summary description of the l probable impacts of the proposed l action on the environment; and :
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(3) The basis for the conclusion that no environmental impact state-ment need be prepared. -
40 C.F.R. 1500.6(e) ,provides in pertinent part:
. . . if an agency decides that an environ-mental statement is not necessary for a pro-posed action . . . (iv) for which the agency has made a negative determination . . ., the agency shall prepare a publicly available record briefly setting forth the agency's decision and the reasons for that determina-tion . . . .
20/
See Consumers Power Company (Midland Plant, Units 1 and 2), ALAB-458, 7 NRC 155, 162-63 (1978).
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be negligibly small. The impacts of any alternative must be equal or greater, and it has been held that " [a] n alter-native which would result in similar or greater harm need not be discussed." Sierra Club v. Morton, 510 F.2d 813, 825 (5th Cir. 1975). We therefore need not consider alternatives 21/.
or the need for 'he t modification in~any detail. Indeed, in the opinion of this Board, not only is such consideration
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unnecessary, it is inadvisable, since it infringes upon prerogatives and duties of corporate management outside our statutory purview. To be sure, were there substantial adverse environmental impacts, our duties under NEPA would require us to balance them against benefits and examine less damaging alternatives. But where, as here, the proposed ,
l action has no such impacts, we can leave considerations such I as economic advantage and capacity requirements to those within the licensee to whom such decisions are normally entrusted.-22/ .
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22/ The Board would, at this point, note that in response I to the NRC's "No' tice of Intent to Prepare Generic En-vironmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel" (40 Fed. Reg. 42901, September 16, 1975), the Staff prepared a draft generic environmental impact statement (NUREG-0404, March 1978) and issued in August 1979 its " Final Generic Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel." Thus it is no longer neces-sary for the Staff in each case to apply, weigh and balance the five factors discussed in the Notice of Intent. Were such consideration still necessary, such factors are described, evaluated and balanced by the Staff in the EIA (Exhibit 6C at 22-26).. While not a ccace 'ed matter subject to Board review, the Board has satisfied itself that the discussion therein meets (Ft. 22/ cont. on next page)
i Colemans' Contentions 2 and 6 l
- 2. The licensee has given inadequate consideration to the occurrence of ac- l cidental criticality due to the increased density or compaction of the spent fuel ;
assemblies. Additional consideration of l criticality is required due to the follow-ing:
A. deterioration of the l neutron absorhtion
[ sic] material provided
. . by the Boral plates lo-cated between the spent !
fuel bundles; B. deterioration of the !
rack structure leading to failure of the rack and consequent dislodging of spent fuel bundles;
- 6. The licensee h'as given inadequate con-sideration to qualification and testing of Boral material in the environment of pro-tracted association with spent nuclear fuel, in order to validate its continued properties for radioactivity control and integrity.
- 16. The Board consolidated consideration of the Colemans' Contentions 2 and 6, and will treat them together in this decision in that they both deal with material property and compatibility consideratIions relative to the new racks for the spent fuel pool.
22/ (continued) the requirements of the Commission as contained in the Notice of Intent. Were the Board required to make a conclusion regarding the five factors, it would be as follows: The Board has applied the ". ve
. factors set forth in the Commission's " Notice or Intent to Prepare Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel," 40 Fed. Reg. 42801 (September 16, 1975), and concludes that they favor issuance of the requested license amendment at this time.
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- 17. The increase in design capacity from the present -
264 fuel assemblies to the new capacity of 1170 fuel as-semblies would be achieved by installing new racks with a decreased spacing between fuel storage positions. The old ra'cks had'a nominal' center-to-center spacing between fuel storage l'ocations of'21 inches. The new racks would be modular stainless steel structures with individual storage cavities to provide a nominai center-to-center spacing of
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10.5 inches. Each stainless steel wall of the individual cavities would contain sheets of Boral, a trade name for l boron carbide in an aluminum matrix, to provide for neutron absorption. The spent fuel pool is located in a separate )
fuel handling building adjacent to the reactor containment 1
building (SER, Exhibit 6B at 1-1, 2-1; Exhibit 1C at 4-6b, l 12-24; Tr. 419).
- 18. Exxon Nuclear Company is the supplier of the racks for the Salem facility. It was responsible.for the design, the engineering analysis and the projecu management and the quality assurance required during the design and construc-tion stage of the racks (Tr. 602). The Licensee has audited 1
the quality assurance program of the Exxon Nuclear Company l 1
1 (Tr. 490-99) and is, of course, ultimately responsible for the quality of the installed racks.
- 19. The racks are designed such that under the worst postulated conditions, i.e., with fresh fuel with no burnable
- poison and a fuel loading of 44.7 grams of uranium 235 l
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isotope per axial centimeter of fuel assembly, the Keff, which is a measure of the approach toward criticality of an array, is equal to or less than 0.95. Keff would have to be equal to or greater than 1.0 for criticality to occur. The 0.95 standard has been se~t to give sufficient margin to
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. : avoid criticality considering calculational techniques and experimental verification . (Tr. 657-59; Exhibit 6B. at 2-1; Exhibit 1C at 11-21). The accuracy of the calculational models utilized to predict the Keff of the spent fuel pool racks have been verified by benchmark calculations by Exxon Nuclear Company for the Licensee and have been approved by the Staff (Tr. 655-56; Exhibit 6B at 2-1 to 2-3). If instead of analyzing the worst case, the typical case of spent fuel discharged to the spent fuel pool were examined, the Keff would be 0.75 or less even without credit for Boron in the water (Tr. 557). Even if an entire Boral plate was missing in a 5 x 5 storage array, the Keff would still meet the 0.95 Keff criteria (Tr. 576).
- 20. No credit was ta' ken for the effect on criticality resulting from the presence of the boron in the spent fuel pool (Tr. 550). During the refueliz.J process, the spent' fuel pool water comes into contact with the water in the reactor cavity, and the spent fuel pool water is borated to prevent its dilution (Tr. 445, 736-38; Exhibit 1C at 22).
The borated water remains in the spent fuel pool even after
refueling has been completed. Even with fresh fuel, the fuel storage racks are designed such that Kaff would be equal to or less than 0.95 without any Boral plates whatscever if credit were taken for the 3oron present in the pool (Tr.
576-77).
- 21. The only materials used in the fuel storage racks, the rack interties,_and wall restraints are Type 304 stain- l
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less steel and the Boral material sealed between an anner :
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- l and outer stainless steel shroud (Exhibit 2, 12; Exhibit 11, i Response to Question 13). The stainless steel shroud pro-tects the Boral from exposure to the spent fuel pool water environment (Exhibit 1C at 3A; Exhibit 6B at 2-12). Type 304 was chosen for its compatibility with the spent fuel l water, which contains boric acid at a nominal concentration of 2000 ppm boron, and is the same material which is utilized in the present spent fuel racks (Exhibit 2, . .12 ; Tr. 736-38). Stainless steel of this type has been,widely utilized in the nuclear industry (Exhibit 2, 112-3,' Tr. 456).
- 22. Based upon experience gained in environments similar to the Salem spent fuel pool, there is no evidence that any corrosion or other deterioration of stainless steel would take place (Exhibit 2, 13; Exhibit 6B at 2-15, Exhibit 7 at 5-6; Testimony of John R. Weeks following Tr. 652 at 2-3 [ hereinafter " Weeks at __"] ; Tr. 670-71) . Stainless steel fixtures have been exposed in pools up to 20 years without evidence of degradation (Tr. 480; Exhibit 2, 13; Exhibit 7 1
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at 10). Boral material has also been exposed in water for periods up to 20 years without any significant deterioration (Weeks at 3 and BNL-NUREG-25582 appended to the testimony; Tr. 662-63). No significant deterioration is expected for the boric acid environment in the spent fuel pool (Weeks at 4).
.23. The. Licensee. has, made. detailed and comprehensive ,
plans to. assure'tha't the fabricated racks are built and installed in accordance with specifications designed to assure their continued ability to perform their intended function (Tr. 495-99). Special procedures will be utilized during handling and installation of the spent fuel racks, including the use of a specially designed crane to assure that the installation complies with all requirements (Tr.
597; Exhibit H (Handling, Shipping and Receiving Inspec-4 4 tion]). One of the design considerations was the loads to which the racks would be subject during all, phases of fabri-cation, shipping,' handling and installatio'n. These loads are significantly lower than other loadings for which the l
racks were designed (Tr. 478-79). As part of the quality assurance effort, careful control of the manufacturing process and nondestructive testing of the fuel cells has been conducted to assure at least 95% leak tightness with a 954 confidence level (Exhibit 2, 15; Exhibit lH; Tr. 458, 492, 616-17, 634).- In addition, quality assurance and process checks have been performed on Boral to be utilized l i
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= l at' Salem to assure that all design requirements are met (;gl camera Tr. 26-55; Exhibit 1C at 23-24).
- 24. The details of the welding processes and other manufacturing and nondestructive and metallographic examina-tion are described in the application (Exhibit 2, 16; Exhibit lH l .. The quality assurance program includes a helium leak
. test utilizing.a helium mass spectrometer which is capable of detecting very small' pin holes, smaller than any which would be significant in the fuel storage cell service en-vironment (Exhibit 2, 16; Exhibit lH). j 1
- 25. In addition to and not withstanding the efforts to l
prevent any leakage, Exxon Nuclear Co., Inc., has conducted a series of experiments to determine the effect of a hypo-thesized leak in the stainless steel shroud. Such a leak could potentially cause some minor corrosion of the aluminum in the aluminum-carbide matrix and the evolvement of hydrogen gar (Tr. 437). The phenomenon would be self-limiting in that the aluminum" oxide film formed at the same time as the hydrogen gas is generated is inert and slows down the cor-rosion rate (;gl camera Tr. 44-45; Tr. 609-13, 691-92; Weeks at 2). The amount of hydrogen gas which could be potentially evolved-is small and would not present any potentially
, explosive mixture at the surface of the spent fuel pool (Tr.
440, 595-96, 612-13). The water leaking in the void between the shrouds would compress the gas at the top of the cell until an equilibirum pressure was reached. The hydrogen gas g- t v. -%_ ._, ,.
,- _ .____m., , , - _ - . - - ,
21 -
would increase the pressure in the gap between shrouds pushing the water level down until gas bubbles escape at the elevation of the crack. The worst location for a leak would thus be at the bottom due to the higher static pressure. If a leak were' hypothesized to occur at the bottom half of the storage cell, the pressure would cause the inner shroud to bulge and. move toward the center of the cell. If the leak
~
had occurred.at the. top half, no bulging would occur (Exhibit 2, 17; Exhibit lH).
- 26. The Exxon Nuclear Company tests revealed that in the unlikely event that a leak in the lower half of the fuel storage cell occurred after installation in the water-filled storage pool and before fuel is inserted, the worst potential consequence would be failure to be able to insert the fuel, j thereby losing the affected cell from service (Exhibit lH l
(Corrosion Swelling Effects); Tr. 579, 605). Prior to i l
loading fuel in any location, a procedure will be utilized to determine whether cell. swelling exists ht that location (Exhibit 2, 18; Exhibit lH [In-Plant Testing Program]; Tr.
580).
- 27. If a leak were to develop in a storage cell with fuel already in place, the most severe result would be that the fuel co'ldu not be withdrawn with the normal withdrawal force of the fuel handling crane (Tr. 605). Exxon Nuclear Co. subjected a dummy fuel assembly to the simulated pres-i ,
l t l l l l
~
l ~
o .
sdre which would result from a hole in the worst location in.
the storage cell. The fuel assembly was instrumented with strain gauges to determine the associated stress. The maximum stress on the fuel pins which the bulging shroud would contact would be 20,000 psi, substantially below the
.yi ld stress of 55,000 psi for dnirradiated Zircaloy-4 tubing .(Exhibit .lH .(Test Report at 1-4] ; Tr. 941, 743, 745-46; Fvhibit II) . -Irradiated zircaloy would. have even a higher yield stress (Tr. 753). The swelling of a fuel cell would also not have any adverse effect on criticality considerations (Tr. 532-53). Semi-remote tooling would be utilized to provide vent holes in the top of the storage cell annulus to relieve the gas pressure on the fuel assembly and permit routine removal (Exhibit lH [ Fuel Bundle Recovery From Bulged Fuel Cell]; Exhibit 2, 19; Tr. 443, 485-89, 715).
The Board concurs that this is an acceptable approach.
- 28. In another series of tests conducted during 1977 and 1978, Exxon Nuclear e;xamined the ability of the Boral to withstand the spent fuel-pool environment under a variety of conditions and considering a range of Boron pres,ent (Tr.
447; In camera Tr. 8, 14- 15) . A number of test coupons of varying configurations, some of which were similar to the storage rack shapes, were exposed to fuel pool type environ-ments for periods of up to one year (p1 camera Tr. 10-13).
l Inasmuch as the performance of Boral is well documented in l
l l
l
. l significantly more severe radioactive environments, it was not necessary to simulate this factor as part of the l l
tests (;gl camera Tr. 13-14,-33-34; Tr. 603-04, Weeks at 4).
The' coupons were examined for corrosion rate, pitting, l
. bonding, edge attack and bulging. Two different phenomena were ,iden'tified (;gl camera Tr. 21-23). One was a. random ~and very localized bulging'which is self-limiting. (;gl camera Tr.
24,'47;' Exhibit 3 and 5 at 3-3, 4-7, 4-8'; Tr. 615-16).
These localized bulges normally occurred in an environment different than will be experienced at Salem Unit 1 (Exhibits 3 and 5 at 4-7). Moreover, even after 30-40 years of ex-posure, the swell would be on the order of a quarter of an inch which is insufficient to affect the clearance between the fuel element and storage location (I_n camera Tr. 22-23; Tr. 693-94). The other phenomenon idvolved was the pre-viously discussed bulging of the storage cells due to oxi-dation (;91 camera Tr. 24). The experiments.showed that simulated storage
- cells with a leak-simula' ting hole will sustain aluminum corrosion which will consume only a small percentage of the aluminum in the Boral core after a 40-year exposure (Weeks at 2). The projections were compared with previous experiments and studies done by others and found to have an acceptable correlation (;gl camera Tr. 39-40).
Moreover, while some pitting, edge attack and internal gas I
pressurization could occur to Boral plates, B 4C particles, l
\
I i
1
. l
which are highly inert in the environment of the spent fuel pool, would not be dislodged in the process and thus no effect on criticality safety would occur IIxhibit 2, 110; Exhibit 8 at 1-5; Exhibits 3 and 5 at 3-2 to 3-5; In camera Tr. 17, Tr. 566,.665-66; Weeks at 2, 4) .
As a result of its expergmental' work and on the basis of previous work, no mechanism has been identified which will degrade the Boral materialfor the pose ,for which it was designed in the
'l' Salem facility (Tr. 435-36) .
- 29. The Licensee, in addition to these test programs, has'connitted to a long term fuel storage cell surveillance j l
program to verify that the spent fuel storage cell retains l the material stability and mechanical integrity over its service life under actual spent fuel pool service conditions (Tr. 497-99; 515-16; Exhibit lH (Long Term Fuel Storage Cell i
Surveillance Program]). Samples of flat plate sandwich coupons and short fuel storage cells will be provided for I
periodic surveillance and testing. The samples are of the same materials and are produced using the same manufacturing and quality assurance procedures specified for the fuel
- storage cells. One short fuel storage cell and one flat i
plate sandwich coupon will be prepared such that the Boral material will be exposed to the spent fuel pool environment (Tr. 585-88). The planned frequency of examination would be about one year after rack replacement and about every two
~
ye'ars thereafter (Exhibit 2, Ull; Exhibit lJ (Long Term Fuel -
Storage Cell Surveillance Program]; Tr. 586-87). The NRC Staff has reviewed the program and determined it is a viable program for keeping track of any corrosion phenomena (Tr.
694-95). The Board concurs. ,
. 30.,
Thera was considerabl'e discussion during the ,
. . course.of the hearing concerning. problems encountered at other operating facilities'regarding these sp'nt e fuel racks.
Initially, the racks at the facilitites discussed were sup-plied by vendors other than Exxon Nuclear Co. Because of this and the differences in design, construction and quality assurance requirements, experience at these other facilities has limited relevance to the issues in this proceeding (Tr.
438-39; Exhibit E, 112; Tr. 458-59, 461). These spent fuel pools in question had not been designed to be leaktight (Tr.
439, 457). In any event, PSE&G and Exxon Nuclear Co. have, by virtue of their quality assurance programs, nondestructive testing, and long-term surveillance programs for the fuel pool, assured that prob 1dms which have occurred at other facilities are not likely to occur at the Salem Generating Station (Rvhibit 2, 112; Tr. 442-43). Moreover, the long-term surveillance programs to be conducted by PSE&G and the experimental programs already conducted by Exxon Nuclear assure that there is no health and safety problem associated with the' fuel pool, even should the spent fuel pool environ-
)
ment come into contact with Boral. The periodic sampling l and testing of the Boral coupons would detect any incipient deterioration. Thus there is no substance to the Colemans' assertions regarding Boral.
~
- 31. 'In reaching this conclusion, the Board has con-
.sidered the relevant portions of a report dated April 10, 1979, made by the Commission's .05fice of Inspection and Enhorcement on.its. findings relating to an inspe~ction conducted from March 19 to March 23, 1979~at the Monticello Nuclear Generating Plant, operated by the Northern States Power Company at Monticello, Minnesota. The report found that after new spent fuel storage racks had been installed in the spent fuel pool at Monticello, 11 of the 676 fuel storage cells would not accept an Oversize go/no-go gauge used to check'the dimension of the cells, and that of these 11, two would not accept a dummy fuel element. The change in the dimensions of the cells appears to have been caused by swelling of the cell walls due to the buil' dup of gas re-leased within the walls by a chemical reaction between water and the Boral material. After the cells had been removed from the pool, vented (by drilling h'o les in the top of the cell walls), resized by vacuum and mechanical means, and reinstalled in the pool, 6 of the original 11 cells would still not accept the go/no-go gauge. All of the cells accepted the dummy fuel element, however.
i
27 -
- 32. The Monticello racks were not designed or manufac-tured by Exxon Nuclear, the supplier of tPr Salem racks aad were of a different design (Tr. 458-59). Each of the af- l facted storage locations was resized "by vacuum and mechanical means," a procedure which has not been proposed for Salem. l As previously discussed at paragraph 26, at Salem, were an
]
empty cell to swell, it would be considered unavailable for - j
~
l use (Tr. 580, 605,.609). There are no plans to return such i
a cell to service. If a fuel cell at Salem which had a fuel assembly stored in it developed a leak, a ccmpletely dif-farent situation would arise. With a fuel cell stored l inside, a fuel storage cell would not bulge beyond its elastic limit and would thus return to its original shape when pressure was remcVed by venting (Tr. 606). l
- 33. Moreover, there are substantial design differences between the spent fuel racks at Salam and Monticello (See ,
for example, Tr. 457-459). One substantial
- difference is that extreme care has been taken to assure that the Salem spent fuel storage cells 'are completely sealed while the Monticello cells were not designed to be leak-tight (Tr.
437, 443, 626).
- 34. Even accepting the fact that 8 of the 676 fuel storage locations at Monticello (i.e., 1.2%) are not usable, which is conservative since all would accept a dummy fuel assembly and presumably an actual fuel element, there ,has 6
_ , , - - h. -. .- , _ _ - , ,,Q..---,. --
been no showing that this has any significance to the present proceeding. The Colemans have failed to provide any basis for their position that there has been a substantial loss in cell availability at Monticello or connnected it in any way to show that'there is any potential for such'a substantial loss at Salem. -
During.the course of.the proceeding, intervenors L '
- 35. ,
. raised the question of ' venting the cells as an alternative to the course proposed by the Licensee, i.e. keeping the cells sealed. The Board has concluded therefore that the Licensee's program to assure cell integrity (Tr. 619'-20, 622-23) is acceptable. It is our further conclusion that the Licensee's approach is the' superior of the two. Our i I
- determination is based upon the assurance of leak tightness l l
previously discussed and upon the ability to vent the racks at some later point in time should the need arise. I
- LACT Contention 1 ,
The Licensee.has not considere'd in sufficient detail passible alternatives to the proposed expansion of the spent fuel pool. Specifically, the Licensee has not established that spent fuel can-not be stored at another reactor site.
Also while the GESMO proceedings have been terminated, it is not clear that the spent fuel could not by some ar-rangement with Allied Chemical Corp.
be stored at the AGNS Plant in Barnwell, South Carolina. Furthermore, the Licensee has not explored nor exhausted the possibilities for disposing of the spent fuel outside of the U.S.A.
~
- 36. The Board has found, supra, paragraphs 10-15, that the present action does not require the preparation of an environmental impact statement and thus does not require the consideration of alternatives. Nonetheless, for-complete-ness of th'e record, the Board has addressed the merits of this~ contention should a reviewing tribunal disagree with this Board's analysis.
As an, initial matter, the Board views the second sentence of the contention as limiting the
- first, applying the principle of statutory construction, expressio unius est exclusic alterius. The result is that the alternatives suggested by LACT during the course of the hearing do not even come within the scope of the contentions.
Similarly, however, for the sake of completeness, the Board has discussed even those suggested at the hearing for the first time. For the reasons discussed below, the Board finds that alternatives to the proposed expansion of the capacity of the Unit 1 spent fuel pool have.been adequately considered by the* Licensee and the NRC Staff and that none is superior to the one selected by the Licensee.
- 37. The question of when the present and expanded racks would be expected to be fully utilized was fully explored during the course of the proceeding inasmuch as this schedule could possibly affect, to some extent, the viability and attractiveness of the various alternatives. It is expected that, after the initial refueling, refueling would occur on an approximately yearly basis (Tr. 793-94). The first e
.-- G.-- , ..n.n, . , - , ,,-.,n,_., , . . . . _ _ . , ,,n._,, , ~ _ _ , . - . , _ , , - . . .
o .
re' fueling outage occurred 22 months after fuel loading at which time forty, instead of the originally planned sixty fuel elements, were replaced (Tr. 793-94). The Licensee plans to discharge 52 spent fuel elements during the next
discharge and.56 in. subsequent annual cycles (Tr. 1104-06).
.Mua plant spent fuel pool would thus be completely full after the' fourth refueling which is scheduled in 1983 (Tr.
102.6', 1030, 1104-6). 'With regard to the discharge o'f spent
~
fuel from the reactor, this represented only about a one year slippage in the date which the spent fuel pool with the present racks would be completely filled (Tr. 1026). The Board recognizes that as a result of the accident at Three Mile Island, the dates for operation of Unit 2 have been set back (Tr. 1031-33). However, while dates for operation may vary, such changes which have occurred to date and which may reasonably be expected to occur in the next few years would not significantly affect the Board's reason $ng concerning its review of altdrnatives (Tr. 1029-40, l'043-45). In this regard, the Board cannot' ignore tha reality that projected dates for this industry are prone to slippage. This is particularly true of the projected date for availability of any government-sponsored independent spent fuel storage facility,' discussed below.
- 38. During the course of the proceeding, a number of hypothetical questions were permitted by the Board.with
regard to projecting the date on which the spent fuel pool j would be full based upon an assumption that subsequent refueling outages would be as long as the first outage (Tr.
805-06). The intervenors have failed to present any evi-dentiary support for their hypothesis, and the Board has N . . ' i given 'no: evidentiary, weight ' to the res'ponse to these questions.
~
See Pacific Gas and Electric Company (Diablo Canyon Nuclear.
Power' Plant, Units Nos.1 and. 2) , ALAB-334, 3 NRC 809, 825 4
(1976).
- 39. It is not practicable to store the spent fuel from Salem Unit 1 at Salem Unit 2 or either unit of the Hope Creek Generating Station. In the case of Salem Unit 2, 23/
since that unit is expected to begin operation shortly-~
and will have an annual discharge of fuel after the first discharge, both fuel pools with existing racks would be full by 1984 even were their capacities to be shared (Exhibit 6C at 16-17; Tr. 1027). During the course of ,the hearing, it was' suggested on cross-examination that th'e expansion of the capacity of Salem Unit 2 spent fuel pool would face less licensing difficulty and thus the alternative of expanding only Salem Unit 2 should be favored. We have no basis for accepting this premise. Even so, the Board does not agree that the mere possibility of fewer procedural obstacles in 2y The Board will take cognizance of the fact that the NRC recently issued an operating license for Salem
, Unit 2 which permits fuel loading and certain low power testing.
e
, ,- , , - , , - - , , - , - - - , - , - - . ..-,.----c- - , - n.-- -,, -n-
O l$ censing Unit 2 is a proper means to assess the difference in environmental impacts between this alternative and the Licensee's proposal. We see no point in pursuing this approach. In any. event, using an expanded pool at Salem Unit 2 in conjunction with the existing racks at Unit 1 would onlf permit storage until 1991.(Tr. 1039-40).
40p Moreover,.the environmental impacts of the extra handling of.irrad$ated spent fuel, such as the dose received
by workers during the transfer, would have to be attributed to this alternative inasmuch as the spent fuel pools for the unics are completely separated and each fuel element would have to be removed from the Unit 1 spent fuel pool and placed in a cask prior to transfer. The environmental impacts of shipping Unit 1 spent fuel to Unit 2 for storage l l
would appear to give greater exposures than those for changing '
the racks in the contaminated state at this time (Tr. 1137, 1148-52). There does not seem to be any countervailing l advantages to this alternative. None has been demonstrated I to us. Moreover, if the Unit 2 racks had to be changed to the greater capacity racks to increase storage at a later I time and Unit 2 fuel had to be shipped to Unit 1 for storage, the doses would be even greater in comparison (Tr. 1138-( 1152). Due to the uncertainty in the availability of an l
Independent Spent Fuel Storage Installation ("ISFSI") by that time (Exhibit 6C; Tr. 1005), such an alternative could
, impact adversely on Un it 2 operation; the Board considers it l
'e as'only a short-term temporary expedient and not a substitute for the proposed acticn.
- 41. With regard to storage of Salem Unit 1 spent fuel at the Hope Creek units, it is unlikely that these units would be sufficiently complete to enable fuel to be stored pri'or to 'tlia'. .unmodi. fied Saleb.i unit being full- (Exhib'it '6C at-
. .y 17-18; Tr.831). Storage at Hope. Creek would involve replacement of the Hop,e" Creek racks with racks capable of holding Salem 1 fuel, further limiting storage capacity for spent fuel generated at enose units. Again, spent fuel would have to be transported to these units and those im-pacts weighed against this alternative.
- 42. Considering that the same problem with spent fuel ,
l
- pool storage is being faced by all utilities, it is unlikely !
that there will be storage space available at any other reactor spent fuel pool. The costs and environmental impacts i
of such storage would be at least comparable to installing new racks at Salem Unit 1 (Exhibit 6C at 18) . Moreover, Licensee's proposal has no adverse environmental impacts 1
associated with the additional. transfer of spent fuel which is associated with this alternative.
- 43. The Allied-General Nuclear Services ("AGNS")
reprocessing plant has not yet been licensed to receive and store spent fuel in the onsite storage pool. AGNS has l
stated that in.no event will the facility be utilized by
~
l l
AGNS for the storage of reactor fuel absent reprocessing (Exhibit 2, 'I21; Exhibit 6C at 14-15). Considering the President's April 7, 1977 statement deferring indefinitely commercial reprocessing and recycling of the plutonium produced in the U.S. nuclear power programs, the storage
- .a c' pici'ty' o'f. 'that ' fac.ility cannot beJ relle'.d ' upon.'.
Y. , . . ,-,, ,
4 4 ~. The NRC had. unde.r review an applic.ation by Exxon
- ~
Nuclear Co. for a' storage pool and reprocessing facility to be located at Oak Ridge, Tennessee. A construction permit has not yet been issued and, in view of the President's announced policy and the termination of that proceeding by the NRC, reliance upon the construction of a storage pool in time for Salem Unit 1 is not prudent (Exhibit 2, 122).
- 45. The fuel storage pool at the Morrir, Illinois i facility is being utilized for General Electric Company owned fuel which had been leased to utilities or for fuel which General Electric had previously contracted to re-process. Other spent fuel is not being stored in the absence of an express cosmitment to do so. There is no such commitment for Salem (Exhibit 6C at 14; Exhibit 2 a.t 123) .
Similarly, the Nuclear Fuel Services facility at West Valley, New York is not accepting additional' spent fuel for storage, even from those reactor facilities with which it had reprocessing contracts (Exhibit 6C at 14; Exhibit 2 at 123).
1 l
J
35 -
- 46. The Federal government has indicated that a spent fuel repository may not be available until at least 1983 or 1984 (Exhibit 6C at 14-16). However, no legislation has been enacted Tuthorizing construction of such a facility (Tr. 838) ; neiti.er has the environmental impact study of 7 such a project been completed (Tr. 838, 980-82).
Further -
more, no site.has' bee,,n selected.(Tr. 844).. .Recently, the Department of Energy indicated that there may not'be space available for transfer of fuel from plants such as Salem in the near future even if DOE constructed a storage facility in the requisite time period (Tr. 1005-08, 1053). First priority would be given to hard-pressed utilities (Tr.
1053). This is borne out by the February 12, 1980 statement of the President's program on waste management which states at page 2 that the administration is still pressing "for legislation to build or acquire limited spent fuel storage capacity at one or more away-from-reactor facilities for those utilities unable to expand their storage capabilities
. . . (emphasis supplied)'. " Thus dependence on space in a government owned ISFSI is not sufficiently firm to constitute a viable alternative to the proposed action.
- 47. Should an ISFSI be constructed, the costs would be much higher than those associated with the new racks for Salem Unit 1 inasmuch as a new pool structure with special design requirements, i.e., some seismic resistance, and
s pporting systems would have to be designed and constructed, and spent fuel transported to such a facility (Tr. 986; Exhibit 2 at 524). Ever d.e witness for LACT has recognized this (Testimony of George Luchak, Ph.D., Transcript follow-ing 918 at 2-4 (het .:.af ter "Luchak at __"] ) . In addition,
~ .
land would have.to be acquife'd,'and it.would be.necessary<to -
overcome licensing problems (Exhibit 2 .at 124) . 'The enviren- .
- J ,
mental impacts associated with constructing su'ch a facility would also be greater than the minor impacts associated wita replacing the racks (Tr. 790, 835, 977-80, 1083-84). The l
proponent of this contention, LACT, presented the testimony l of Dr. George Luchak. Taking his testimony as admitted by the Board as a whole, it adds essentially nothing to the statement of the contention. At most, the implication is left that for some unspecified accident, the location at a dry, unpopulated site would, in some unquantified way, be better. However, Dr. Luchak has apparently only studied a basic nuclear en'gineering textbook (Tr. 895-96) , has not 1
written any scientific articles concerning this subject j matter (Tr. 896-97) , has never visited the Salem facility or any other nuclear power plant for that matter (Tr. 897) , has l never~ studied the layout of the plant and was not acqua'inted with the detailed design of the plant or fuel pool (Tr.
900). Furthermore, the Board 1,as found Dr. Luchak to be totally unqualified to perform any accident or radiological
as'sessment evaluation (Tr. 894, 907-09, 913-14). In any event, this witness could not postulate any particular accident directly affecting the spent fuel pool; only vague allegations about indirect effects from some serious, but unspecified reactor accidents involving meltdowns could be
- ; 'given. (.Tr. -8 5 2,. 9 54-5 5, 9 8 7) -. - The Board has determined that e .- -
in.so(ar as,certain , accidents were not includable in the design bas'is by th'e Commission at th'e operating license
~
stage, their postulated impacts need not be considered as part of the review of alternatives (Tr. 1047-50). In any l event, Dr. Luchak is unqualified to provide any expert opinion regarding accident probabilities, consequences or :
I risks and has given his testimony no weight. The Board finds '
that an ISFSI, particularly one in a distant desert site, i 1
not to be a superior alternative.
- 48. All previously discussed alternatives assumed that the spent fuel pool could be filled priot to the alternative being needed. This is not quite the case. After the next (second) refueling, the facility will lose its capacity to ,
discharge a full core from the reactor. While this capability is not a safety-related consideration, it is prudent from an operational standpoint to retain such capability (Tr. 866-69). Therefore, the ability to sustain full core discharge capability should be weighed in favor of the proposed fuel
! rack replacement.
l l
, . - . -. - . a,
4
- 49. The Licensee has discounted the possibility for disposing of the spent fuel outside the United States.
Considering the President's announced policy statement on l nuclear power, it is unlikely that permission would be granted to export spent nuclear fuel. In fact, the President's
, . April 7, .1977 statement on nucleab power policy[ states that -
the U.S. is explori.ng ,"measur_es.to assure access to nuclear ,,
~
fuel supplies an'd'sp'ent Nuel storage (in the United States]
for nations sharing common non-proliferation objectives."
More recently, as part of his program on radioactive waste management, of which this Board has taken official note, the President h's a stated that the U.S. would accept "linited amounts of foreign spent fuel when the objectives of the U.S. nonproliferation policy would be furthered." Thus, the Board finds that disposal of Salem spent fuel outside the United States is not a viable alternative.
- 50. Inasmuch as the Commission has at the operating license stage, as*part of its environmental review, made a determination that there was a need for power and while not raised by any intervenor or participant, the " alternative" l 1
of shutting down the facility may not be considered by this Board. See Dairyland Power Cooperative, LBP-80-2, supra, 11 NRC at 65-74. However, for the sake of completeness and l
should a reviewing body disagree, we have included a dis- l cussion of this matter. The Licensee has estimated that a ,l r
l
. 39 -
shutdown of Salem Unit 1 with a net electrical output of 1090 megawatts would cause incremental replacement power costs alone of $500,000 per day, based on the differential costs.of producing energy.from Salem as compared to produc-tion from other available units in the PSE&G and Pennsylvania Interconnection (E'xhibit' 2. at',
New . Je'rsey , Mary, land "("PJM i") .
, ;
127). The Staff, .looking. at the long term economic. impacts j
-l other than 'the short term incremental effects, 'factiored in 'a. .i 1
capacity factor range of 60-70% to arrive at annual replace- l ment costs associated with the discontinuance of operation on the order of $300,000 to $350,000 per day (Exhibit 6C at 18-19). Using either figure, these costs would still be far i 1
in excess of the costs associated with the proposed modifica- l tion, i.e., $3,300 per fuel assembly or $3,000,000 for the entire cost of replacing the racks (Exhibit 6C at 19) .
- 51. During the course of the hearing, the Board inquired as to the criteria to be utilized in deciding whether to ship spent fuel offsite if the application to expand the spent fuel pool were gran'ted and offsite storage became available at a later time (Tr. 869-72). The primary con-sideration would appear to be an economic one inasmuch as the other factors which might come into play, i.e._, environ-mental and safety, do not appear at this time to be de-terminative (Tr. 873-8). The Board believes it appropriate l
to leave such a determination to the Licensee, based upon
th'e information which is . available to it at the time a l
decision in this regard is to be made, such as the availa-bility of permanent disposal facilities.
- 52. In conclusion, should the Board be~ required to review alternatives to the proposed action it finds that
. other 'avai'lable ' alternatives 'are not preferable to the installation of the increased cap'a city racks.
It therefore ,
finds that-LACT Contention 1 has no merit.
, BOARD QUESTIONS
Background
- 53. During the course of this proceeding, the accident at the Three Mile Island Nuclear Station, Unit 2 ("TMI-2")
occurred. As a result, the Board posed three questions regarding the impact of that accident on the TMI-2 spent
.. fuel pool, and the effect on the spent fuel pool if a "TMI-2" or " meltdown" accident were transposed to the Salem Unit 1 facility: -
- 1. To what extent did the accident at Three Mile, Island affect the spent fuel pool at that site?
- 2. If there had been an explosion or i
" meltdown" at Three Mile Island, l what effect would that have had upon the spent fuel pool? To what extent would it have mattered how much spent fuel was present at the pool?
- 3. If an accident such as the one at Three Mile Island occurred at Salem, to what extent would the accident affect the spent fuel pool? If an -
explosion or " meltdown" occurred at
-+
s Salem, to what extent would that affect the spent fuel pool? To what extent would it have mattered how much spent fuel was present at the pool at Salem? LBP-80-10, 11 NRC 337, 342 (February 22, 1980).
- 54. In response to the Staff's objection, we withdrew
.. our.sec.on'd question as. unnecessary to this proceeding and
.s. -
postponed the time,sh'en'we'wou.ld hear svidence on the other l
', questions.
The Staff, joined by'the Licensee, al..soobje[cted to' that portion of the Board's third question concerning the effects of an explosion or meltdown on the Salem spent fuel pool, asserting that the question impermissibly required consideration of Class 9 accidents. The Board heard evidence on its first question and the unchallenged portions of its third question on July 11, 1979.
- 55. With regard to its Question 1 and the unchallenged portions of Question 3, as may be seen from the statement of its questions, the Board was not attempting its own general review of the TMI,2 incident; the only purpose of these questions was to identify, the possibility of a safety con-cern arising because of the changed design of the racks as a result of the hypothesized occurrence of a TMI-type accident at Salem Unit 1. As discussed, infra, this Board is satisfied, after questioning the witnesses presented by the Licensee and Staff and after considering the matters raised on cross-examination by the parties, that no safety problem associated with the installation and use of the increased capacity
shentfuelrackshasbeenidentifiedandtheexistingdesign bases are adequate in this regard.
- 56. At the July 10, 1979 hearing session, the Board posited a fourth question to the parties, asking, in effect, whether the TMI acciden't was a Class 9 accident:
~
Th'e proposedfAnnex'to Appendix D, 10 ,
, CFR 'Part 150, appears to define a Class
, 9 accident as a sequence of failures ,
which are.'more severe than.those whi'ch '
I the safety features of the plant are " '
design ~ed to prevent. The sequence of failures at Three Mile Island pro-duced a breach of the containment and a release of radiation which could not be prevented by the safety features.
Was the occurrence at Three Mile Island therefore a Class 9 accident? Was the risk to health and safety and the en-vironment " remote in probability," or
" extremely low" at Three Mile Island, !
as those terms are used in the Annex?
(Tr. 922-23).
- 57. On February 22, 1980, after receiving the parties' l
l varying responses to its fourth question, including the l Staff's answer characterizing TMI as a Class 9 accident, a position which was not free from dissent 5 rom within, the Board issued its Februar? 22, 1980 Memorandum and Order ad-dressing, inter alia, the Licensee's and Staff's objection to its previously posed third question. The Board discussed recent developments concerning the authority of adjudicatory boards to consider the consequences of Class 9 accidents, focusing on the Appeal Board and Commission opinions in Offshore Power Systems, (Floating Nuclear Power Plants),
-O ALAB-489, 8 NRC 194 (1978), aff'd. on certification, CLI 9, 10 NRC 257 (1979).
> 58. In view of the responses to Question 4 in our Memorandum and Order, we posed a fif th question to the parties for. consideration at an evidentiary session:
. "In ' he' t event of a gross loss of water from the storage pool, what would be the difference in consequences between
- those occasioned by the pool with ex-
. panded storage and those occasioned by the present pool (11 NRC at 346) ?
- 59. In our Memorandum and Order, 11 NRC at 345, we male an analogy between a floating nuclear plant and the Salem Station since Salem is located on an " artificial island" and discussed our tentative views on the treatment of Class 9 accidents as the result of the TMI accident. The Licensee subsequently asked the Appeal Board to consider whether the Lice'nsing Board's fifth question was proper. This request was denied as premature in Public Service Electric and Gas Company (Salem Nuclear Generating Station, Unit 1) , ALAB-588, 11 NRC __, (April 1,' 19 8 0) . Because there may have been some confusion concerning our views due to the subse-quent issuance of a number of decisions dealing with this matter, and as the result of the March 28-30 evidentiary l l
session which permitted this Board to directly apply the generalized treatment of Class 9 accidents to the factual 1
context of this proceeding, a further explanation and ;
an'alysis of the Board's position is appropriate.
i
- 60. On March 21, 1980, the Commission issued its Memo- !
24/
randum and Order in the Black Fox proceeding; in finding that the Appeal Board--25/had misinterpreted its decision in ;
26/ '
Offshore Power,- the Commission vacated that portion of ALAB-573' which directed the Staff to file its views on
.I
.iwhethTr C ass 9 ac'cidents.should'be considered'in t'he
- l 2
Black Fox. proceeding._,7,/ .
]
- ;
l61. The Commissi6n. reiterated that'the existing policy '
against consideration of Class 9 accidents for land-based plants was not changed by Offshore Power and stated its belief that its policy on consideration of Class 9 accidents for land-based plants would not properly be developad by rulings on a case-by-case basis because " piecemeal considera- l
-28/ '
tion is not appropriate to such an important policy area." - )
l It added:
I
' Because the existing policy on Class 9 accidents was not displaced in Off- l shore Power and would not be displaced
. 1 24/ Public Service Company of Oklahoma (Black Fox Station, Units 1 and 2) , CLI-80-8, 11 NRC (March 21, 1980).
See also ALAB-587, 11 NRC (Mar E 28, 1980), which gave the Licensing Board lidtructions consistent with the Commission's decision. I 2_5,/ Public Service Company of Oklahoma (Black Fox Station, Units 1 and 2) , ALAB-5 73, 10 NRC 775 (December 7,19 79) . I 26/ . Offshore Power Systems (Floating Nuclear Power Plants) ,
CLI-79-9, 10 NRC 257 (September 14, 1979).
22/ CLI-80-8 at 1-2.
28/ Id. at 3.
e--,- - , - - - . - , -
_r- -- -rwv-- -
---w --m -r,e -v-e=w ww-v- - --
e- e-
. pending generic consideration of Class 9 accident situations in policy develop-ment and rulemaking, the Co= mission en-visioned that the staff would bring an i individual case to the Commission for decision only when the staff believed that such consideration was necessary or appropriate pri'or to policy develop-ment. 29/ ,
. .. 'l 62.
Th'us,.the Commission has ' reserved to its' elf the
, decision of whether class 9 accidents may be consi,dered in ,.
- any given case. 29/ ' . Absent affiMmative action by the Com--
mission, the Staff and licensing boards are prohibited from proceeding with any such consideration.
- 63. The Commission's Black Fox decision also specifies that the recommendation to the Commission must come from the Staff, which has an affirmative duty to bring to the Com-mission those cases which it believes warrant consideration i.
--31/
of Class 9 accidents. We interpret the Commission's Black Fox decision to mean that the Appeal Board, and by l implication, licensing boards, are not empowered to order
^
the Staff to make.such an affirmative recommendation to the 32 /
Commission. The Appeal Board found this " unambiguous ex-pression of Commission policy" controlling in this Salem 3
proceeding. 1l -
2_9/ I_d_. at 3-4 (emphasis supplied) . See also id n.'3.
3,0/
0 ALAB-588, 11 NRC at _ (slip op. at 9).
31/ Id.
3_2/ Id. at _ (slip op. at 9-10)./ Id.
- 64. As part of its response to Question 5, the Staff presented its views with regard to whether, in accordance with the criteria set by the Commission in Black Fox, 34/
-~
supra, Class 9 accidents should be considered in..this 35/ proceeding.-- It stated that "[t]he expansion of the spent
' fuel pool at the ' Salem' site does not constitute an exceptional
, case.... . resulting in risks.substantially greater than for
~
an average plant" and 'its ultimate conclusion that "the environmental consequer.ces of Class 9 accidents need not be 36/ evaluated."-- ! =
- 65. In its February 22, 1980 Memorandum and Order, the Board drew an analogy between a floating nuclear plant and the artificial island site upon which the Salem reactors are founded with regard to the potential for liquid pathway releases were a Class 9 accident to be postulate.d. As a 34/ The Board and Staff witnesses have utilized a defini-tion of Class 9 accidents consistent .with that noted in ALAB-588,'n.2, 11 NRC at __ (slip op. at 3). See Tr. 1464 and the Board's Memcrandum and o-der, LBP-80-10, 11 NRC at 342-43.
35/ The Board recognizes tha't the Staff's silence would have been sufficient to preclude consideration of Class 9 accidents and that under ALAB-587, Black Fox, the Board could not have ordered the Staff to present its views. ' 36/ Direct Testimony of Walter F. Pasedag in Response to Board Question No. 5, following Tr. 1 87'at 5 (herein-after "Pasedag at "). See also Tr. 1469. The Board notes the Board Notification in this docket dated April 24, 1980 transmitting, inter alia, a decision of the Director of Nuclear Reactor Regulation under 10 C.F.R. 52.206 which at pp. 28-39 addresses in detail why the Staff does not believe it appropriate to con-sider Class 9 accidents for this facility.
, 'N
result of the evidence before it, the Board concludes that it is clearly not appropriate to consider Saler as having a liquid pathway comparable to a floating nuclear plant; rather such liquid pathway is comparable tu that of the land-based plant used as.a basis for comparison in the
~
.'LiiNiid Pathway ,' Generic Study .(NUREG- 440-) .
~
J Groundwater ! transport, surface water transport, and usage of the . water-bodies' surroundi'ng the . Salem' site' were examined (Pasedag 'at 3). The Staff evaluation indicated slower dispersion of postulated releases via the liquid pathway compared to the typical estuary site (Pasedag at 3). In fact, the Salem I site was used as the model for a land-based plant in the NUREG-0440 study, with the exception that the fish popula-tion from the Chesapeake Bay was utilized to produce mere conservative results .(Tr. 1600-04). Thus, the Board is able co conclude that special consideration of Class 9 accidents i at the Salem site is not necessary because of any unique considerations associated with the liquid pathway.--37/
- 66. The Commission's Statenent of Interim Policy, p.
11, supra, does not require consideration of additional . matters in this proceeding. Initially such statement does not require any further action in this proceeding as the record was closed prior to its promulgation and prior to the 37 / We would note that were there a reason for closer scrutiny of 'he t liquid pathway, none of the other parties or participants addressed this as a signifi- l cant pathway resulting from the " gross loss of water" in their proferred testimony.
48 - July 1,1980dateafterwh(chadditionaldiscussionmay.be i 1 needed in an impac: statement not cc=pleted by that date. i l Moreover, since no EIS is necessary, discussion of postulated 38/
-~
accidents is not required. The Commission has specifically stated that, absent a showing of special circumstances,
. . . . ;. .
~ \
the Statement should not se.rve ,as a b' asis' for opening,- "
, reopening or , expanding any previous or ongoing. proceeding.
. 'Moreover,'the Commission'expresse'd confidence'th'at; even l
if the revised treatment were to be followed, similar con-clusions regarding the environmental risks of accidents 39/ would be reached by a continuation of current practices.~~
- 67. The Board's next consideration was whether the
" gross loss of water" postulated by Question 5 was an ac-cident that could be classified as one which should be in-cluded within the design basis for this facility. The Staff's testimony analyzed the design of the spent fuel pool in terms of the hypothesis of a " gross loss,of water" advanced by the Bdard. The Staff found th't a the high density racks have no appreciable effect on the structural stability and seismic response of the spent fuel handling building (Pasedag at 1-2) . It concluded that the leak tightness of the expanded pool under all postulated accident conditions is assured and no appreciable change in the margin of pro-taction arises from the pool modification (Pasedag at 2; Tr.
1463-64). . 38/ See p. 11, sup ra . 39,/ Statement of Interim Policy , 45 Fed. Reg. at 40103. 1 - - - - - .Q, -,_.y - - . - --
l 1
. 68. The Staff testimony analyzed the weld channel leak '
detection system in terms of an initiating mechanism for a l l
" gross loss of water." The Staff explored a scenario which 40/ l involved multiple punctures of the fuel pool liner l
~
leading ' to a maximum leak rate of no more than 710 GPM cal-culated on a conservative ~ basis and probably substantially
. less, which,would be. detected and appropriate action taken
. i .
to preclude uncovering of any of the ' fuel' (Pasedig at 2; Tr.- 41/
-~
1461-62). Inasmuch as any additional release of radioactive material to the spent fuel pool water due to the presence of any additional stored fuel is insignificant, and the amount of water lost is independent of the fuel racks being utilized, the differences in radiological consequences of such a spill of water would be insignificant (Pasedag at 2).
- 69. While the Board did not take evidence as to the )
l causation of a " gross loss of water," sne believe it profit- l able to review the initiating causes as advanced in the testimony responsive to Question 5 to set 'the consequences in perspective and to view it in terms of other risks as-40/ The Licensee's testimony, Exhibit 14 for identification, was similar in this regard. See also Tr. 1470. 1 41/ On cross-examination, the Staff witness stated that ! while this event involved multiple successive failures, he did not consider this to be a class 9 event because it would not cause an accident which would uncover the stored fuel (Tr.1461) . The Board believes that under the definition that it is utilizing, even this event constitutes a Class 9 accident because of the assumed successive failures. However, no mechanism has been established for causing multiple liner punctures.
a ciated with the facility. As explained previously, the Board has not felt it necessary to re-review the design basis of the facility for its protection against natural phenomena. None of the parties have advanced any reason for doing so. Dr. Wdbb, a witness for Lower Alloways Creek Township,'has advanced ,the idea of an earthquake larger than
, the, NRC has required for.the design basis as causing ."a
~The' Board' believes'that such a hy-gross loss of water."
pothesis without foundation is a nullity and need not be further considered. The Board also does not believe that it need consider the consequences of sabotage in view of the l 42/ NRC's requirements contained in 10 C.F.R. 573.55. 4 Neither does it believe in view of the findings it has made with regard to criticality and Boral, that it need consider an initiating event deriving from some hypothetical unde-fined criticality. The final causative agent for a " gross loss of water" advanced by Lower Alloways Creek Township is some vague, but serious reactor accident which would not 2 directly affect the spent fuel pool, but would somehow make access to it and its auxiliary components significantly more difficult or would be so severe as to require all personnel to leave the site for an unspecified (but long) period during which the spent fuel pool is unattended and a key, but unspecified, component associated with the cooling system is assumed to fail.
'M Commonwealth Edison comoany (Zion Station, Units 1 and 2), LBP-80-7, 11 NRC 245, 283-85 (February 14, 1980).
1
- 70. ~With regard to reactor accidents causing an impact -
on the spent fuel pool, the Board has examined only one
- specific scenario in some detail, the events associated with the TMI-2 accident. The results of this evaluation are 1
discussed, infra. With regard to the present discussion, suffice it to say that such accident would not have had a
.significant effect on.the spent fuel pool; it would take an
~
~
" accident significantly more severe to affect the spent fuel pool in the manner suggested by LACT's witness, Dr. Webb.
In effect, Dr. Webb assumes that such an accident occurs and then makes additional assumptions of equipment failure to analyze in isolation the effect on the spent fuel pool. This approach so distorts any perspective on a systematic, coordinated safety or environmental review as to be a mean-ingless method for judging the safety design basis of a particular system or f acility or to serve as the basis for j disclosure of environmental impacts. . i ] 71. Having given further review to the responses of the l l parties and participants'to the Board's fourth question I related to whether the TMI-2 events were a Class 9 accident, the Board is of the view that even an affirmative answer
'has, in the final analysis, limited utility in determining whether a'" gross loss of water" need be analyzed. If we had authority to consider them, it would be an oversimplification to lump all Class 9 accidents together. The fact that a
specific event occurred may have only marginal significance wich regard to accidents which have the potential of ad-versely affecting the spent fuel pool. Thus, while this Board has not undertaken a major effort to attempt to place 1 a quantitative value on the probability of an accident !
~
serious enough to' affect' the spen't fuel pool, th'e Board
' ~
. believes that. such an accident is extremely improbable when
~- - ,
' viewed'in'the'spe'c'trum of' accidents'tha't have been previously placed in th'e Class 9 accident category. In fact, the Board has failed to identify any reasonably probable series of events leading to an instantaneous " gross loss of water."
Thus, were we inclined to read the Commission's Statement of Interim Policy on accident analysis under NEPA as requiring any additional consideration in this proceeding, which we explained, supra, at 10-12, we are not, we would nonetheless find the risk associated with a " gross loss of water" to be extremely low. Specifically, the consequences of such a
" gross loss of water" are bounded by the accident consequences discussed in WASH-1400 (Further Testimony of Walter F.
Pasedag in Response to Board Question No. 5, Tr. following 1387 at 3 (hereinafter"yasedagFurtherTestimonyat__,"). This combined with the predicted low probability of occur-rence of the " gross loss of water" permits the conclusion 43/ As discussed above, the Board has not identified any specific sequence which could cause an instantaneous
" gross loss of water." Moreover because of the ad-ditional series of low probability events necessary (Ft. 43/ cont. on next page)
?
thht the risk associated with such occurrences would not, in the Board's view, constitute a significant environmental risk.
- 72. In senmary, this Board believes that the initiating events which could cause a gross loss of water are cumula-Jti elp so remote as to reduce the risk to such an extent as
,s . .to requir.e no;further discussion.
However, in an abundance *
't of cautiori, E.'he*' Board discuss'es 'the evid~ence before it J
relating to the consequences of a hypothesized " gross loss of water." Consequences Of An Instantaneous Gross Loss Of Water In The Spent Fuel Pool . 73. During the March 28-30, 1980 session of the pro-l ceeding, the Board took evidence on the subject of the con-sequences of an instantaneous " gross loss of water" without attempting to define any sequence of events which could lead to that result. The Board wishes to state its conclusions at the outset and,will explain the development of its rea-soning below. The question of the consequences of the postulated event is not free from controversy as it is based upon a series of complicated analyses and assumptions. Taking into consideration the evidence of record, including testimony of the witnesses and giving appropriate weight to each of them, as discussed below, the Bo.$rd finds that 43/ (continued) for a reactor accident analyzed in WASH-1400 to affect accessibility to the fuel pool for extended periods, this event would have a probability of occurrence much lower than the WASH-1400 reactor accident itself.
. - . - _ _ _ . . _ , . . . ~ . -----a v- --
cohsequences and the overall risk associated with the storage of the additional spent fuel in the new racks (i.e., fuel not presently authorized to be stored by the present operating license, that is, fuel four years and older) compared to the
' storage of' fuel'in the present racks is not substantially greator.
k!/ .The - ' range of consequences and probability has been. developed in the. record o'f this proceeding.
~
The lead wit; ness was Walter Pasedag.. The Board was 74. impressed with the breadth and depth of his knowledge in the area under investigation. The Board was particularly impressed with his ability to meld analytical data with practical experience and applications. For these reasons, the Board was able to place broad reliance on Mr. Pasedag's testimony. Supporting Mr. Pasedag was Dr. Allen S. Benjamin of Sandia Laboratories. Dr. Benjamin was one of the authors of NUREG/CR-0649, Spent Fuel Heatup Following Loss of Water During Storage ("Sandia Report") , a generic . study which sup-plied analytical methodology and calculational techniques 45/ applicable to the Salem case.-~ Dr. Benjamin was fully 5 44/ The risk of a gross ~~1oss'of water has been previously discussod; when both these risk factors are combined, the requested approval presents an extremely small en-vironmental risk. 45/ The Board would note here the following p'a ragraph that was included in the Sandia Report, presumably to place it in perspective: The likelihood of a severe spent fuel pool drainage accident is judged to be extremely low. Many spent fuel pools are constructed below grade, essentially precluding complete (Ft. 45,/ cont. on next page)
- o. _ __ _ __
ap. prised of the analytic techniques utilized in the heatup analysis. The Board found his testimony generally to be cogent and well thought out. Assuming that the most recently ~ discharged rods had reached the temperature for self-sustaining oxidation,'Dr. Benjamin was unable to conclude whether suff'iciAnt h,edt would be 1ransferred to spent fuel rods
, which ha'd be'en stored four years and long~r.to, e permit these rods'tocbeach'.the sel'f-s'usta'ning-oxidation temperature.
~
i This step was not part of the Sandia Report which only 1 examined the status of the most recently discharged batch of l l fuel (Tr. 1433, 1441). Dr. Benjamin stated that he was l unable to reach a conclusion on this question based upon the 45/ (continued) drainage of the pool due to structural , failure. Numerous design features are in-l corporated in all facilities to minimize the likelihood of a loss of pool water, includir.g (1) the conservative design philosophy of building the concrete struc-ture, racks, cooling system, and support structures to withstand the forces that might result from a large earthquake or tornado, (2) de' sign of the racks to as-sure that the geometry of stored spent fuel is maintained in a subcritical con-figuration, (3) location of pool penetra-tions to prevent draining or siphoning of water through associated piping systems, (4) inclusion of mechanical interlocks and operating procedures to prevent the crane from passing over the pool with heavy loads, and (5) provision of multiple water level, water temperature, and radioactivity monitors which actuate alarms in the control room. Stringent security measures are enforced to prevent sabotage. A complete drainage of a spent fuel pool, therefore, has to be considered as an extremely unlikely occurrence (footnote omitted] [Sandia Report at 12] .
-, --m- ,---w-- - - - - - - ,r . - --- --, - w
analytic work he had done to date,--46/ but in his opinion, further investigation was warranted (Tr. 1436-39). While from the standpoint of understanding the basic physical processes involved, such an analysis might be nteresting on
~
a generic basis','the Beard concludes, as discussed below, that the fu'therr analysis cannot be' justified in light of - the evidence which-has.been. received in this decket and . .
~'
c'on'sideding W fh.ndin'gs'tliat this Board is required ti[ make
~ ~ '
f . under the Notice of Hearing and Commission rules and precedents.
- 75. In contrast with the reliance on the testimony of I
- -
these two Staff witnesses, the Board has given little weight to the testimony of the witness sponsored by Lower - Alloways Creek Township, Dr. Richard Webb, based to a large extent upon his demeanor on the stand and his presentation in his written testimony (See, for example, Tr. 1705-06, 1775). Dr. Webb was argumentative and nonresponsive in answering the questions of the parties and the Board (Tr. 1704, 1717-20). In addition, his testimony had negligible
, l
- 46/ At first, in response to a series of Board questions, Dr. Benjamin had expressed a tentative conclusion that such propagation was more probable than not, but after reflection, changed this position to the one discussed in the text (Tr. 1437). The Board is appreciative of his candor and believes that his original answers were made in an effort to be helpful and responsive to the Board; however, the Board acknowledges his final posi-tion made after additional reflection and considering the additional time he devoted to this matter (Tr.
1437) as respecting his best analysis under the cir-cumstances, considering the complicated nature of the physical process involved (Tr. 1438). r .~ ,- . ,. , , , . . . _ .- ~.m .._.7,-n_, , , , ,,- ~ . .
p obative value with regard to the Board's fif th question. Dr. Webb characterized many physical phenomena as " con-ceivable" or "may be conceivable" in that it was physically possible that they could occur or that they could not be ruled out (Tr. 1723-1730, 1769) but he did not attempt to
. ' attach any probability lto each such'eVenti nor'das he able -
to give any such estimate ~from the stand (Tr. 1710-11). s .- -
.- Tluxs f it is not possible to glean from. Dr. Webb's . testimony - .
whether a postulated occurrence had'a probability of 0.1, 10 -7 or 10-14 and, as a result, his testimony was little aid to the Board (Tr.1732) . For example, Dr. Webb would assume that all the cesium and strontium was released into the environment (Tr. 1702, 1731-2). In contrast, he admitted that for the worst reactor accident, WASH-1400 stated that no more than-10% of the strontium would be released (Tr. 1773-74). Furthermore, when challenged on cross-examination to pinpoint the basis of statements which he stated were contained and discussed in his testimony,,he first indicated that he was unable to remember the contents of his own testimony, then claimed that the matter was discussed only by " implication" (Tr. 1704-10) or scattered throughout an entire section. While the Board permitted Dr. Webb to testify on a number of broad areas, the Board is convinced that Dr. Webb does not have the qualifications of an expert l in these fields and the Board has discounted his testimony l heavily (Tr. 1701, 1741-43). Dr. Webb b'ased certain of his l
~
m._- -n -- -. . -- _ L _.- _. .a.
cdnclusions on " experiments" he did with material removed from flash bulbs; the Board does not. consider this to be a true scientific experiment (Tr. 1754). Finally, Dr. Webb
- was admittedly unable to reach any conclusion regarding the ability'of the present racics to sustain the fuel without
. reaching a.self-sustaining oxidation temperature and thus to !
. l testify r,egarding'the' difference in consequences (Tr 1716; ,
)
~
i Webb P'a' r t IiI" 6i suppiement St '1-3 ; and. Open ' Rack Analy' sis of Fuel Heatup, attached thereto). A detailed discussion relating to our own conclusion follows.
- 76. The Board has adopted the Staff definition of a
" gross loss of water" as a hypothetical non-mechanistic, instantaneous loss of all cooling water in the present and expanded spent fuel pool combined with an inability for unspecified reasons, to refill the pool, or providing any other mode of cooling other than natural convection air cooling (Pasedag at 3) . The Staff indicates that this is an incredible event (Pasedag at 3 ; Tr. 1575-7S) . Extending the Staff's reasoning to its, logical conclusion, inasmuch as the probability of this event is extremely small, the associated risk would also be extremely small, although the '
consequences can be finite. The Board is satisfied that there has been no mechanism identified for causing an in-stantaneous " gross loss of water" associated with the in-stallation and use of the higher capacity racks, nor has
I
-m-,+n n ------- y - - - - - ,,,-
a e- - -- , -, ,v,-r,,- ,-,.e ,
1 any scenario postulated for the racks been shown to have 1 an increased probability of. occurrence. No design deficiency 1 l nor need fsr improvement has been identified (Tr. 1604-10).
- 77. For fuel freshly discharged from the reactor, the assumed continued denial of water cooling capability may eventually lead' t'o oxidation and failure of the clad and to overheating of UO2. fuel,fwith the pot.ential for the release
' of' ' fission products in the UO'2 ' fuel in either the presen't or the expanded ~ pool (Pasedag at 4; Tr. 1441). The doses at the site boundary resulting from this postulated release would depend heavily on the postulated scenario for the
! mechanism of the water loss, subsequent cooling attempts and building integrity (Pasedag at 4; Tr. 1442). The witness for the Staff, .'r. Pasedag testified that even where the clad was damaged to a point where it lost its integrity, either by complete oxidation, or by partial oxidation and melting, and considering the physical processes involved and I the design of the building, there would on'ly be a small release of activity, and.it would not be significant at the site boundary (Tr. 1445-47). Even were fuel melting to be
- postulated for the one-third of the core, i.e., the latest batch-to be discharged, then the consequences would be similar to those postulated for a reactor accident in WASH-1400 except the consequences would be somewhat less than one-third as much because only the equivalent of one-third g . 6 .. + . - , - - , , . , - - - - , .-'
w,--.c e ,-. .---gL - - - - -
;
of a core is involved and there has been some decay time 47/ since operation of the core at full power (Tr. 1447-48). Looking at the older fuel, while there is a possibility of some fission product release from older fuel, it would only be a few isotopes and would be small (Tr. 1448-50, 1452,
.~
1526-27, 1600-01). 'The Board recognizes that the comparison
...to.a-reactor, accident is not exact', but.still considers that
~
it[db$s' haves'ubshanEi16t'iliSy. 'It' demons'trates tha't the' environmental consequences associated with the assumed
" gross loss of water" are not unbounded nor would they be significantly beyond those previously discussed by the NRC.
In fact, the WASH-1400 cases appear to bound the consequences of the event at issue (Pasedag Further Testimony at 3) . In its prefiled testimony, the Staff utilized the onset of self-sustaining clad oxidation as a conservative criterion for the release of the fission products of the fuel in order to estimate the differences in the potential consecuences of this hypothetical event arising from the p' col modification 48/ (Pasedag at 4) .~- The new storage configuration may result 47/
-- Furtbormore, somewhat less than one-third of the core is scheduled to be discharged from the reactor each year.
48/
~~
It is worthy of note that the situation at Salem Unit 1 with regard to the onset of self-sustaining clad oxidation is better than for other facilities having expanded racks; the free volume above the pool is significantly larger than for the typical plant (Tr. 1400-01, 1596-97), and the downcomer space at the side of the pool is greater than for certain other plants having expanded racks (Tr. 1590-91). i l t
, , . - , _ - -. ., w A -, , .- .n--.. U ,...:-.
-.,.q--,,.
in less natural convection and hence a higher likelihood of reaching oxidation tamperatures and possible clad melting for recently, discharged fuel.--49/ Although heating of fuel assemblies. stored adjacent to the most recently discharged assemblies would occur, it is t e view of'the Staff, which is being adopted by the~ Board, that there has been no credible mechanism,forlthe propagation of a " zirconium fire" or the spreading of other than limited oxidation to the four year old or older fuel stored in the pool as a result of its expansion (Pasedag Further Testimony at 2; Tr. 1393-94, 1396-97, 1412, 1442-44).
- 78. On cross-examination, Dr. Benj amin stated that his analysis for Salem was based upon the worst configuration with regard to the position of the spent fuel elements in relation to each other (Tr. 1406, 1445; See also Tr. 1452-53). For-example, he testified that if a " checkerboard array" were utilized, the newly discharged batch would not reach a self-sustdining oxidation temperature after 60 days of discharge (Tr. 1454, 1572-73). This array could be maintained until half the capacity of the pool were utilized without any shuffling of fuel and through the penultimate discharge if fuel were shuffled (Tr. 1573, 1585-86). Even 49/ Dr. Webb states that while he has not been able to do i a rigorous analysis he does not believe that the open !
frame racks are "much more coolable than closed racks," citing WASH-1400 (Webb's Part III' dated April 20, 1980 at 1-2). Thus there must remain even some doubt as to this conclusion. j 1
4 co.nfiguration may result in less natural convection and hence a higher likelihood of reaching oxidation temperatures 49/ and possible clad melting for recently discharged fuel. Although heating of fuel assemblies stored adjacent to the most recently discharged assemblies would occur, it is the view. ol' the Sta'ff, whicl$ is being' adopied by the Board' that' there has,been,no credible mechanism for the propagation'of l a! " zirconium.: fire" or the spreadi'ng of'o'ther than limited . oxidation to the four year old or older- fuel stored in the l pool as a result of its expansion (Pasedag Further Testimony at 2; Tr. 1393-94, 1396-97, 1412, 1442-44). 1
- 78. On cross-examination, Dr. Benj amin stated that his analysis for Salem was based upon the worst configuration ,
l 1 with regard to the position of the spent fuel elements in relation to each other (Tr. 1406, 1445; See also Tr. 1452-53). For example, he testified that if a " checkerboard array" were utilized, the newly discharged batch would not reach a self-sustaining oxidation temperature after 60 days of discharge (Tr. 1454, 1572-73). This array could be maintained until half the capacity of the pool were utilized l without any shuffling of fuel and through the penultimate discharge if fuel were shuffled (Tr. 1573, 1585-86). Even 49/ Dr. Webb states that while he has not been able to do a rigorous analysis he does not believe that the open
. frame racks are "much more coolable than closed racks,"
citing WASH-1400 (Webb's Part III dated April 20, 1980 at 1-2). Thus there must remain even some doubt as to this conclusion. l
- _* -m.,-p- . _ . , a
_,_y. . , . .r
O we're fuel in the old racks assumed not to melt immediately 50/ after discharge should a " gross less of water" occur, there would only be a relatively small period of time involved during which even the latest fuel discharged to the new racks would approach the self-sustaining oxidation temperature (Tr. 1457, 1497, 1503-05, 1508-09). If the newest elements-
~
do.not reach.the self-sustaining clad ~ oxidation temperature, then the' older elements also will not (Tr. 1456). Thus, considering all the evidence of record and the associated uncertainties, the Board does not believe that there is any significant difference in the risk associated with the use of old and new racks, even considering the postulated " gross loss of water" (Tr. 1416-18, 1579-80, 1583, 1599-1600). The Board's analysis is borne out by the position of witness Pas 9.dag that, while small, the risks associated with the movement of fuel would be higher than the risk in not keeping the checkerboard configuration described above (Tr. 1585, 1587). Thus, the ' Board does not believe that there is even any reason for prescribin'g the location of spent fuel as-semblies in the pool.
- 79. The Board is unable to pinpoint a credible mechanism by which a hypothetical, self-sustaining clad oxidation pro-pagates from a newly discharged spent fuel element to an old 50/ Mr. Pasedag estimates that with the present racks using existing ventillation, the fuel would be cool-able in air in on.the order of 10 days (Tr. 1457).
If . . - o , ,. _
spent fuel element. The zirconium fuel rods postulated to heat up would form a substantial zirconium oxide layer which would inhibit the oxidation (Pasedag Further Testimony at 2; Tr. 1734-35). The mechanism whereby the other assemblies were raised to their self-sustaining oxidation temperatures assumed by Dr. Webb in his testimony was that of a " zirconium fire," a deflagration with columns of. visible flames shoot- I in'g up from'the spent fuel pool with'large convection forces', air currents and drafts (Webb's Addendum at 2-3; Webb's Part I at 31-35). When challenged, Dr. Webb retreated frcm his original definition of " zirconium fire" (Tr. 1752-53) but did not otherwise define a. specific mechanism for the spread of the self-sustaining oxidation. To achieve the large frac-tional or total releases predicted by Dr. Webb would, as a minimum, require a physical mechanism such as the " zirconium fire" described above. To a large extent, Dr. Webb based his assumption of a possible deflagration on a fire which occurred on April *28, 1955 at the Bettis A'tomic Power Labora-tory in contiguous open bins of oil-contaminated zirconium scrap (Tr. 1751, 1770). The scrap consisted of 92,000 pounds of turnings and chips, 12,000 pounds of solid zir-conium scrap and 55,000 pounds of miscellaneous zirconium
. scrap. The report of the investigating team concluded that the solid scrap did not burn, but most of the other scrap did (Tr. 1514, 1591-96).
5 .-
. j l
- 80. The pyrophoric properties of zirconium in a finely divided state are well known. The witnesses for the Staff testified that af ter Dr. siebb had raised the spectre of a
" zirconium fire" they had investigated the literature and i
spoken with recognized experts in the field. Both concluded
~
that the Bettis fire experience was not applicable to the situation. at. hand because of the difference in the state of the zirconium '(Tr. 1518, 1521-22, 1535, 1537, 1565-68). .; only in its finely divided state does zirconium appear susceptible to a fire as postulated by Dr. Webb (Tr. 1524-25). The Staff witness Pasedag attributed the initial flareup at Bettis to the contaminant present (Tr. 1511-12). The Staff witnesses relied on experiments of fuel element heatup in an air environment and practical experience in the handling of zirconium rods to demonstrate that the zirconium fire would not occur (Tr. 1497, 1509, 1526-29, 1537-40, 1543-44). .
- 81. Dr. Webb*tried to counter this by reference to the literature in which a recognized expert, Dr. Baker, had ad-dressed the pyrophoric properties of " aggregates," which he defined as including lathe turnings, shavings or powder (Tr. I 1517). Dr. Webb extended the definition of aggregates l l
1
- beyond that contemplated by Dr. Baker to include fuel as-senblies (Tr. 1744-45, 1749-50', 1788-89, 1742-93, 1799-1800). It is clear to the Board from the context of the
55 - paper that this was an improper application and entirely unwarranted. To find that Dr. Baker's article supported his hypothesis, Dr. Webb apparently deliberately ignored other portions which were counter to his position. For example, he ignored Dr. Baker's statement that single pieces of
. .. m. .
zirconium heated. slowly do not ignite even'if heated to
. . 1
~
, 1300* C becau.se of.the tough. oxide film that developed (Tr.
' 1734 35, 1740). Dr. Webb also'ign' o red'in his' calculations the diffusion limitation through the oxide layer (Tr. 1738). !
- 82. Dr. Webb's testimony regarding other postulated physical phenomena associated with the " gross loss of water" also had significant deficiencies. To prove that the integrity of the Fuel Handling Building would be lost, Dr. Webb produced a calculation to show that it would be pressurized to over 100 psi. While Dr. Webb apparently sought to leave the impression that such pressurization was extremely rapid, the methodology utilized had no time element (Tr. 1763, 1766).
It is the conclusion of the Board that Dr.'Webb's assumption of instantaneous pressurization is unwarranted; even assuming the pressure increase took place at all, it would undoubtedly occur more slowly, causing leakage fraa the building to occur instead of a catastrophic rupture (Tr. 1403-04). As a final matter, Dr. Webb's fission product release fraction (essen-tially 100% is released from the building) ignores well known physical phenomena, e.g., plateopt and retention within the fuel and Fuel Handling Building. e 9 . - - -,-,,e-- ---<-m. v
66 - 1
- l
- 83. It is therefore our ultimate conclusion that,
)
1
~ with regard to the " gross loss of water" postulated by the '
Board, no further consideration is necessary in this pro-ceeding. We turn now to a discussion of the TMI accident and 'its postulated effect on the Salem Unit 1 spent fuel 1 pool, the subject of portions.of Board questions l'and 3. i
. . The Three Mile Island Accident-and Its Effect on. '
the Salem. Unit 1 Spent' Fuel Pool Were That' Accident ' Postulated'to' Occur at the Salem Unit l' Reactor
' ~
- 84. With regard to the Board's first question, the record revealed that the spent fuel pool at TMI-2 contained no spent fuel and was empty inasmuch as the unit was in its first cycle ("NRC Staff Reponse In Part To Board Questions, following Tr. 1133 at 2 (hereinafter " Staff TMI Testimony at
__,"]) ; Licensee's Response to Licensing Board's Question 1 and Part 1 of Question 3 Relating to Impact of a Three Mile Island Type Incident on the Salem Unit 1 Spent Fuel Pool" at 2 (hereinafter " Licensee's TMI Testimony at.__"). Even had there been fuel stored in the pool, there would have been no effect (Staff TMI Testimony at 2; Tr. 1235-36, 1272). :
- 85. The Board's next inquiry, i.e., the first part of question 3, was directed towards the effect on the Salem j fuel pool if the TMI sequence of events were to have occurred l I
at Salem. Based upon published data (Staff TMI Testimony at )
- 3) , the following sequence o'f events was postulated to have occurred. After the reactor scram at TMI-2, which was Q , , ,
~-
m _ - ' ' ". - - - . - , .
caused by a loss of feedwater to the steam generators and a turbine trip, a series of events occurred.which resulted in damage to fuel assemblies in the reactor core. A relief valve on-the reactor coolant system pressurizer opened . during the initial pressure transient and failed to reseat, resulting in an overflow of reactor cools.nt system water
.from.the. reactor coolant drain tank to the reactor building. .
'.('c ontainment) sump'. .Thereactbr~ build 3ngsumppdmpsstarted automatical'ly due to the rising water level and discharged water into tanks located in the auxiliary building. These ;
1 tanks became full and overflowed into the auxiliary building. I l Because this water was contaminated from contact with the damaged fuel in the core, the resulting radiation levels in the auxiliary building were high (Staff TMI Testimony at 3) . l
. Based on the information on the TMI incident available to it, the Licensee postulated that the accident resulted in the release of radioactive fission products. contained in the fuel rod gaps (void spaces between the ura'ium n dioxide fuel pellets and the zircaloy cladding) in the reactor core due to clad failure. In addition, radioactive noble gases in l the fuel pellets were apparently released from the core resulting in a total estimated release of approximately 30 percent of the core radioactive noble gases. The activity was released to the prim'ary reactor cooling water, a portion
- of which was spilled onto the reactor containment floor.
l Some of the gaseous activity in the released reactor coolant, l
+
pdrticularly noble gases, became airborne in the containment structure. As noted above, some of the spilled reactor coolant was pumped into the auxiliary building liquid waste storage tanks, and a portion of this water spilled onto the auxiliary building floor, resulting in the radiation levels
'in the auxiliary building and substantially all of the radia-tion levels in the surrounding site environs (Licensee's TMI
' ~
' Testimony at 1-2) .
- 86. An analysis was performed by postulating a release of 30 percent of the gaseous radioactivity frem the Salem Unit 1 reactor core into the reactor coolant system. It was further assumed that all of noble gases released to the reactor coolant became airborne in the containment building.
Finally, it was postulated that radioactivity would be re-leased into the auxiliary building in the area of the rad-waste system storage tanks (Licensee's TMI Testimony at 2-3; Tr. 1265).
- 87. It must again be stated that this Board's purpose is not to determine whether a TMI-type accident could occur at Salem. There are design differences which would preclude a TMI-type accident from occurring at Salem. For example, the containment isolation valves in the transfer lines from the Salem Unit 1 containment sump are automatically shut on a safeguards signal which starts the safety injection pumps.
These valves at TMI-2 remained open during the initial
;
l i i
siages of the accident because they were designed to close only on high containment pressure (4 psig), not on the safeguards signal. This pressure in containment did not reach those levels until abour 20 minutes into the accident, during which time contaminated water was transferred to the auxiliary building liquid rad waste storage tanks as dis-cussed earlier. Therefore, it would not be expected that the. automatic transfer of th contaninated water. in the containment sump would occur at Salen Unit 1 as it did occur j 1 at TMI-2 (Staff TMI Testimony at 3-4; Licensee's TMI Testimony l l at 2-3; Tr. 1268, 1274-75 i. I
- 88. While, because of these and other differences.in design cetween the TMI units and the Salem units, the postulated series of events could not occur at Salem Station as they did at TMI, for analysis purposes, they were assumed to occur (Tr. 1185). Moreover, the physical design of the various components associated with the spent fuel pool and l
their locations are different. For exampl'e, at Salem the spent fuel pool is locatgd in a separate fuel handling 1 building rather than in the auxiliary building as at TMI-2 (Staff Testimony at 4; Tr.1268) . Even in this postulated set of circumstances, there would be no significant impact in the Salem Fuel Handling Building where the fuel storage pool is located or on the spent fuel pool itself. The Fuel Handling Building walls in the area adjacent to the reactor t _ - - .. ~ - -. .. --
co,ntainment are 6 ft. thick at elevations up to 30 ft. above grade (this corresponds to elevation 130 which is the operating
. deck) and 2 ft. thick at higher elevations. The reactor containment walls are 4 1/2 ft. thick. The maximum radiation levels in the Fuel Handling Building from the direct radia-tion in the primary containment were calculated by Licensee te be .less than 400. mrem / hour. Even this radiation level, which would occur only in one small area, would not preclude access to the Fuel Handling Building (Licensee's TMI Testimony at 3).
- 89. Components of the spent fuel pool cooling system that are not located in the Fuel Handling Building are located in areas of the auxiliary building that are separated by at least several feet of concrete from the liquid or waste storage tanks or other equipment which became contaminated as a result of the TMI accident. Hence, the radioactivity in those tanks or compartments would not prevent access to spent fuel pool egoling ' system components, although protective measures such as a radiation work permit or breathing apparatus may be necessary (Tr. 1320-22, 1335). The analysis was based upon actual survey measurements at Three Mile Island to determine the activity of specific components (Staff 330: Testimony at 5 ; Tr. 1181, 13 33-34) . The Salem Unit 1 auxiliary building vent _.ation system is designed to prevent the movement of airborne radioactivity from one e
, 7- jy ,g w A %" e mN #- M >M v* - ~ - - m 2SA.Ad
potentially contaminated area of the building to another. Thus, any gaseous activity released from spills in the liquid radwaste system would not contaminate areas associated with spent fuel pool cooling system components (Licensee's TMI Testimony at 3; Tr. 1292, 1278-79). 90.' 'The ' fans and' filtering ~ equipment .ih the Fuel
~
Handling Building ventilation system are located in the ,
. . t .
' penetration a'rda at elevation 100 ft. (i.e., ground ~ level)'
~
between the containment and auxiliary building. Since this equipment is operated remotely from the Control Rocm, access to this equipment is normally not required. In any event, maximum radiation 1evels, which are calculated to be approxi-
;
mately 12 rem / hour would not be high enough to preclude 1 access. If necessary, dose levels could be significantly -
;
I reduced by placing temporary shielding in front of the l personnel access hatch in the penetration area, the prinary 1 contributor to radiation level in the area. , A single row of lead bricks would reduce ,the radiation levb1 to less than 10 mrem /hcur (Licensee's TMI. Testimony at 4; Tr. 1288-89). I
- 91. Both the additional heat load and the radioactivity in the spent fuel pool as a result of the expansion are not significant (Exhibit 6B at 52.2; Exhibit 6C at 55.3).
Therefore, the amount of spent fuel stored in the spent fuel pool at Salem is not considered to be important to the consequences of the hypothetical accident discussed above. 1
O Moreover, the proposed activity under review by this Board-would not significantly affect the analysis. Thus, regard-less of the absolute significance of the postulated event at Salam, it is ineensitive to the amount of fuel stored at Salem. This fact alone would satisfy the Board's limited inquiry (Licensee's TMI Testimony at 4; Staff TMI Testimony
. at 6). ,, ,
- 92. 'However, based upon the record before us, we can go further. It can be concluded that a TMI-type incident would not result in any significant adverse impact on the storage of spent fuel in the Salem spent fuel pool were the increased capacity racks in place. In fact, it would be negligible (Licensee'c TMI Testimony at 1-4; Staff's TMI Testimony at 6; Tr. 1322). l CONCLUSIONS OF LAW
- 93. The Licensing Board has thoroughly reviewed and evaluated the evidence submitted by all parties with respect to the contentione raised by the intervenors herein which are issues in this proceeding and with respect to the mat-ters raised by the Board. The Licensing Board has also considered all of the proposed findings of fact and conclu-sions of law submitted by the parties and participants.
Based upon its evaluation of the Staff's Safety Evaluation and Environmental Impact Appraisal, the Licensee's applica-tien, as amended, the written testimony of all of the
o witnesses, the exhibits in evidence, as well as the . answers elicited from these witnesses in response to questions of the Board.and the parties and participants, the Board makes the following conclusions of law: (1) That there is reasonable assurance that the activities authorized by the operating license amendment can
.be conducted without endangering the health and safety of the public;.
~
(2)' 'That the activities authorized by the operating license amendment will be conducted in compliance with the Cecmission's regulations; (3) That.the issuance of the operating license amendment will not be inim-ical to the common defense and security; (4) That the issuance of the license amendment is not a major Commission action significantly affecting the quality of the human environment and that it does not require the i preparation of an environmental in-pact statement under the National Environmental Policy Act of 1969,
; as amended, 42 U.S.C. 4321, et sea.,
and Part 51 of the CommissioFs regulations, 10 CFR Part 51' and therefore that alternatives to the proposed action need not be con- l sidered. ORDER
- 94. Wherefore, it is ORDERED, in accordance with the Atomic Energy Act, as amended, and the regulations of the Nuclear Regulatory Commission, and based on the findings and conclusions set forth hereia, that the Director of Nuclear Reactor Regulation is autherized to make appropriate find-ings in accordance with the Commission's regulations and to g-- -,. --~c-,--- ,--, <
-- - . - ,s-, ,. - - . . - - . - - - - .
_ 74 -
~
s idsue the appropriate license amendment authorizing the requested expansion of the spent fuel storage pool capacity at the Salem Nuclear Generating Station, Unit 1.
- 95. It is further ORDERED, in accordance with 10 C.F.R.
S52.760, 2.762, 2.764, 2.785 and 2.786, that this Initial Decision shall be effective immediately and.shall constitute the final action of the Commission forty-five (45) days after the issuance thereof, subject to any review pursuant to the above cited Rules of Practice. Exceptions to this Initial Decision may be filed within ten (10) days after service of this Initial Decision. A brief in support of the exceptions shall be filed within thirty (30) days thereafter (forty (40) days in the case of the NRC Staff) . Within thirty (30) days of the filing and service of the brief of the Appellant (forty (40) days in the case of the NRC Staff) , any other party may file a brief in support of, or in oppo-sition to, the exceptions. IT IS SO ORDERED. - '
,THE ATOMIC SAFETY AND LICENSING BOARD
Frederick J. Shon, Member Dr. James C. Lamb, III, Member Gary L. Milhollin, Esq., Chairman Dated at Bethesda, Maryland this day of July, 1980. g -- - - -
, APPENDIX A - LIST OF EXEIBITS Exhibit No. Description Transcript (Identified, Received) 1-A . Letter from Librizzi to 358- 368 Lear, 11/18/77 -
1-B Letter from Librizzi to 358 368 Lear, 12/13/77 - 1-C Letter from Librizzi to 358 368 Lear, with revised applica-tion, 2/14/78 ' l-D Letter from Librizzi to 358 368 Lear, with enclosure, 5/17/78 1-E Letter from Librizzi to 358 368 Schwencer, with enclosure, 7/31/78 1-F Letter from Librizzi to 358 368 Schwencer, with enclosure, 8/22/78 1-G Lctter from Librizzi to 33 8 368 Schwencer*, with enclosure, 10/13/78 , 1-H Letter from Librizzi to 358 368 Schwencer, with enclosure, 10/31/78 1-I Letter from Librizzi to 358 368 Schwencer, with enclosure, 11/20/78 1-J Letter from Librizzi to 358 368 Schwencer, with enclosure, 12/22/78 1-K Letter from Librizzi to 358 368 Schwencer, with enclosure, 1/4/79
, _, , , _ _ _ _ _ , .- - - - - ~
- e
, 'o ,
Exhibit No. Description Transcriot (Identified, Received) 2 Affidavit of Edwin A.. 358 368 Liden, 2/21/79 3 Exxon Nuclear XN-NS-TP-009 359 413 4 Request for protection of 360 ~414 proprietary information and affidavit in support thereof by Exxon,'Inc'.' 5 Exxon Nuclear 360 414 XN-NS-TP-009 NP 6-A Letter from Schwencer 364 369 to Librizzi, 1/15/79 6-B Safety Evaluation Report 364 369 6-C Environmental Impact 364 369 Appraisal 7 Report of John R. Weeks, 365 652 Corrosion of Materials in Spent Fuel Storage Pools, 7/77 8 Report of John R. Weeks, 367 652 Corrosion Considerations
- in the Use of Boral in
- Spent Fuel Storage Pool Racks, 1/79 ,
9 Letter from Cunningham to 398 Smith, 12/20/77 10 Letter from Crockett to 399 Beckjord, with enclosure, 1/19/78 11 Affidavit of Thomas G. 940 941 Eckhart, 6/18/79 6 g a . .- 3 _ _ . -g - . _ . ,
e 3-
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Exhibit No. Description Transcript (Identified, Received 12 General Arrangement of 1338 1338 Auxiliary Building 13 Monticello Inspection 11 NRC at 339 Report, 4/10/79
~14 ' Lic'ensee's Response to - 1652 Licensing Board Question 5' Regarding A " Gross Loss of
' Water" from ,the Salem Spent
- 3
" l Fuel Pool I (unnumbered) Statement of Estimated Dose Submitted with Licensee's from Moving and Storing Response to Motion for Spent Fuel frcm Salem Unit Reconsideration of Cole-No. 1 in the Unit No. 2 Spent Fuel Pool and State-man's Contention No.
Thirteen ment of Fuel Elements Shipped from Salem Unit No. 1 to Unit No. 2 (Attach-monts to Affidavit of Robert P. Douglas)
. 1 9
G i 9
e
,..) .
. UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSII:: )
Before the .?.tomic Safety and Licansin? Ecard In the Matter'of )
)
PUBLIC SERVICE ELECTRIC AND GAS ) Docket No. 50-272 COMPANY, et al. ) (Proposed Issuance of
) Amendment to Facility (Salem Nuclear Generating ") Operating' License' Station, Unit 1) ) No. DP R- 70 )
CERTIFICATE OF SERVICE I hereby certify that copies of " Licensee's Proposed Findings of Fact and Conclusions of Law in the Form of An Initial Decision," dated June 13, 1980, in the captioned matter, have been served upon the following by deposit in the United States mail this 13th day of June, 1980: Richard S. Salzman, Chairman Mr. Frederick J. Shon Atomic Safety and Licensing Member, Atomic Safety and Appeal Board Licensing Board Panel U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission' Washington, D.C. 20555 Washington, D.C. 20555 Dr. W. Reed Johnson Dr. James C. Lamb, III Member, Atomic Safety and Member, Atcmic Safety and Licensing Appeal Board Licensing -Board Panel U.S. Nuclear Regulatory 313 Woodhaven Road i Commission , Chapel Hill, N.C. 27514 l Washington, D.C. 20555 l Chairman, Atomic Safety and Mr. Thcmas S. Moore. Licensing Appeal Board Panel Member, Atomic Safety and U.S. Nuclear Regulatory l Licensing' Appeal Board Commission U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 Chairman, Atomic Safety and 1 Licensing Board Panel 1 Gary L. Milhollin, Esq. U.S. Nuclear Regulatory Chairman,-Atomic Safety Commission and Licensing Board Washington, D.C. 20555 1815 Jefferson Street ' Madison, Wisconsin 53711 Janice Moore, Esq. Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 v - - -- , .. .
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, f.9 RichardHluchan,Esq. Mr. Alfred C. Coleman, Jr.
Deputy' Attorney General Mrs. Eleanor G. Coleman Department cf Law and 35 "K" Drire Public Safety Pennsville, New Jersey 08070 Envircr;r. ental Protection Section Carl Valore, Jr., Esq. 36 West State Street valore, McAllister, Aron Trenton, N.J. 08625 & Westmoreland
- Mainland Professional Plaza Richard Fryling, Jr., Esq. P. O. Box 175 As.sistant Genefal Solicitor Northfield, N.J. 08225 Public~ Service Electric.
and Gas Company Office of the Secretary
'80 Park Place;. * '
. ' Docketing #and Service Section Newark,.N.J. *07101 U.S. Nuclear' Regulatory Commission l Raymond E. Makul, Esq.
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Washington, D.C. 20555 Assistant Deputy Public Advocate June D..MacArtor, Esq. Division of Rate Counsel Deputy Attorney General ' 10 Commerce Court Tatnall Building Newark, N.J. 07102 P. O. Box 1401 Dover, Delaware 19901 Sandra T. Ayres, Esq. Department of the Public Advocate 520 East State Street Trenton, N.J. 08625 h *
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