ML18082A578

Forwards Licensee Proposed Findings of Fact & Conclusions of Law in Form of Initial Decision.Licensee Reserves Right to File Reply Brief After Svc of Proposed Findings & Conclusions of Law by Other Parties

ML18082A578
Person / Time
Site:	Salem
Issue date:	06/13/1980
From:	Wetterhahn M CONNER, MOORE & CORBER
To:	John Lamb, Milhollin G, Shon F Atomic Safety and Licensing Board Panel
Shared Package
ML18082A579	List: ML18082A578 ML19318A145
References
NUDOCS 8006180509
Download: ML18082A578 (82)

Text

LAW OF.FICES CONNER & MOORE 1747 PENN.SYLVANIA.. A.VENUE. N. W.

TROY :B. CONNER, JR. W A.SfilNG'rON, D. C. 20006 ARCH A. HOOP..E, .JlL

• HABX ;J. WETJ.'ElUIAllN ROBERT H.. RADER June 13, 1980 Gary L. Milhollin, Esq. Mr. Frederick J. Shon Chairman, Atomic Safety Member, Atomic Safety and and Licensing Board Licensing Board Panel 1815 Jefferson Street . U.S. Nuclear Regulatory Madison, Wisconsin 53711 Commission Washington, D.C. 20555 Dr. James C. Lamb, III Member, Atomic Safety and Licensing Board Panel 313 Woodhaven Road Chapel Hill, North Carolina 27514 In the Matter of Public Service Electric and Gas Company, et al.

(Salem Nuclear Generating Station, Unit 1)

Docket No.* 50-272

. Gentlemen:

Enclosed is a copy of "Licensee's Proposed Findings of

• Fact and Conclusions of Law in the Form of an Initial Decision." Pursuant to 10 C.F.R. §2.754(a) (3), Licensee reserves the right to file a reply brief after service of

• proposed findings and conclusions of law by ~he other parties.

Sincerely, Mark* J. Wetterhahn Counsel for Public Service Electric and Gas Company, et al.

MJW:sdd Enclosure cc: Per Service List soo&1soS"o't G

UNITED STATES OF AMERICA

,, NUGLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Mat tsr o,f )

)

PUBLIC SERVICE, *ELECTRIC AND ) Docket No. 50-272 GAS COMPANY, et al. ) (Proposed Issuance

) .* of Amendment to FaciJ.i ty .

(Salem Nuclear Generating ) Operating License Station, Unit l) ) No. DPR-70)

• LICENSEE' S.

• PROPOSED FINDINGS* OF FACT .AND

• CONCLUSIONS OF LAW IN THE FORM OF AJ.~ INITIAL DECISION Public Service Electric and Gas Company, et al.,

Licensee in the captioned proceeding, in accordance with 10 C.F.R. §2.7~4 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 initiai decision.

Respectfully submitted, CONNER & MOORE

.. Mark J. Wetterhahn Counsel for the Licensee Of Counsel:

Richard Fryling, Jr., Esq.

Assistant General Solicitor Public Service Electric and

.Gas Company SO*Park Plaza, TSE Newark, New Jersey. 07101 June 13, 1980 800 618.0 513

... G

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION J

ATOMIC SAFETY AND'LICENSING BOARD Gary L. Milhollin, Esq., Chairman Mr. Frederick J. Shon, Member Dr. James C. Lamb*,* III, Member

• Ih the Matter of 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)

July I 1980 APPEARANCES MARK J. WETTERHAHN, Esq., of Conner & Moore, Washington,

o. c., and ,,.

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.. Alf red C. Coleman, J~.

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 Public Safety, for the State of New Jersey.

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.

INITIAL DECISION (AMENDMENT TO OPERATING LICENSE)

PRELIMINARY STATEMENT l.. This initial decision involves the application ..

filed on November 18, 1977, with the Nuclear Regulatory Commis.sion ( "NRC") by Public Service Electric and Gas Company (PSE&G". or "Licensee") for itself and* as agent for the other. owners,.Atlantic City Electric Company, Delmarva.

• _Po~er *a~d**~ight- Company;* and Phiiadelphia* Elect~i°c* Company, for amendment of Facility Operating License -No. DPR-70 for Salem Nuclear Generating Station, Unit No. 1 ("Salem Unit l" or "facility") located in Salem County, New Je-rsey. The amendment would revise the provisions of the Technical Specifications, Appendix A to Facility Operating License DPR-70, to permit the substitution of new spent fuel storage racks to-*increase the fuel storage capacity from 264 to 1170

• fuel assemblies in the spent fuel pool of the facility.

2. On February 8, 1978, the NRC published in the Federal Register (43 Fed., Reg. 5443) a notice of "Proposed Issuance of Amendment to Facility 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

• . _]:/

the proceedings. The Board in this proceeding has been

_y See Establishment of Atomic Safety and Licensing Board to Rule on Petitions dated March 16, 1978 and Notice of Hearing on Amendment of Facility Operating License dated April 26, 1978.

_:y reconstituted on two occasions. After a prehearing con~ -

ference held on May 18, 1978, the'Atomic Safety and Licensing Board admitted two intervenors, Lower Alloways Creek Township

("LACT") and Mr. and. Mrs. Alfred c .* Coleman, Jr. ("Colemans")

' 3/ .

as *parties.- Requests to participate pursuant to 10 C.F.R *

. §2.715(c) *were received from the States of New Jersey* and 4/

Delaware- and_ wer_e granted _by the ~oar~.- A,11. c;:>r .. part_ of.

• Cc:>J:itentici'ns. 2~*

• 6,. 9 and. i3 *of *the,'Colema.ns

• 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

. Appraisal ("EIA") to the Board and *parties~* Pursuant to the Board's Order Following Specia~ Prehearing *conference dated May 24, 1978, discovery in this proceeding which had begun 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 sunnnary disposition Qf the contentions filed by LACT and the Colemans. The Board .. granted the motion as to LACT

. 5/

Contention 3 and the. Colemans' Contentions 9 and 13 *.~

_y Notices of Reconstitution of Board dated March 8, 1979 and June 27, 1979.

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.

_!/' ~ - Memorandum and*Order dated.Api:il 26, :1978 at-15 .and Order Following Special Prehearing Conference dated May 24, 1978 at 2.

Order dated April 30, 1979. One member of the Board would have granted summary disposition on all issues

. in the proceeding.

5. The record of this proceeding consists of the transcripts of the prehearing conferences held on May 18, 1978 (Tr. 1-120), March 15, 1979 (Tr. 121-316)i 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

. . *.* 7/ ...

discuss information propr1etary.to ..

one of* Licensee's

. *, .. vendors-and the exhibits which were re*ceived. in evidence listed in

_:y Appendix A hereto. Numerous limited appearance statements.

were received. In response to a Board request, the Staff responded directly to a number of the questions raised and 9/

statements made;-* *the Board is satisfied that the Staff has

~/ 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.

...11 The transcript of the in camera session has been received by the Licen~ee and Exxon Nuclear Company and only three pages were ultimately determined to involve proprietary

• matters (Tr. 704-05). The Board has determined that, along with Exhibit 3~,

• the designated pages, In Camera Tr. 22, 36 and 59, should be withheld from public dis-closure.

As a resuJ.t of Board questions and comments, certain information was.submitted by the NRC Staff and Licensee regarding occupational doses *. No objection to the re-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 on Radioactive Waste Management" issued on February 12, 1980. .

See, for example, the July 26, 1979, letter from Staff counsel Janice Moore to Mr. Marvin Lewis.

responded appropr.ia tely. 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 11 scenario. These matters are addres*sed,

• infra *

. .* _:* __<6.~

• _'A.ny __ proposed *findings o*f 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 unnec*essary 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_~ranted, by the NRC in 1976. This Board has recognized the limited nature of this proceeding and has comported itself accordi~gly. The Boarq believes it is necessary to_give some exposition as to its views regarding its role as defined by the Commission. This Board has recognized through this proceeding that it has not been established to conduct a re-review of the decisions reached duri~g the Commission's operating license review, from either a safety or an environmental standpoint, or, as a

co.rollary, 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 rev.iew 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 Eiectric Company (T*roj an. Nuclear _Plant) ,.*

ALAB-531, 9 NRC 263, 266, n.6 (1979) and Northern States Power Company (Prairie Island Nuclear Generating Plant, Units land 2), ALAB-455,*7 NRC 41, 46, n.4 (1978), remanded 2!!_ other grounds, sub~* New England Coalition on Nuclear 10/

Pollution v.* NRC, 602 F.2d 412 (D.C. Cir. 1979) . - We also 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 tliese systems. Of course, this Board is fuzther limited to deciding the contested 11/

issues before i t . -

10/ *Cf.Dairyland Power Cooperative (La Crosse Boiling Water Reactor), LBP-80-2, 11 NRC 44 (1980).

See Detroit Edison Company (Enrico Fermi Atomic Power Plant, Unit 2), LBP-78-10, 7 NRC 381, 386 (1978). The Board is, however, mindful of 10 C.F.R. §2.760a and has posed a nUIIlber of questions, including some related to TMI-2.

~~*-'...--,

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 12/

unchanged by approval of the requested actiqn .* 7 Furthermore**,**:.

iz{ .reviewing. certain. asp'edts of this proposed change, the

.incremental impact was* examined*,

• rather than attempting to.*

. e~tablish an abso.lute . basis for licensing as was done by the*

• *

• Commission at the ..operating license. s1:age.

• This is in accord with the directive of* the Appeal Board in Prairie Island that a board need only consider whether the amendment itself wou-ld 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 whether* alternatives to the proposed action need be considered at all. The National Environmental Policy Act of 1969,

§102 (2) (c) , 42 U.S. C. §4332 (2) (c) (NEPA) prdvides in pertinent part that "all agencies o'f the Federal.Government shall -

~

• * * (C) include in every recommendation or report on pro-posals for legislation and other major Federal actions sig-nificantly affecting the quality of the human environment, a detailed statement by the responsible official on - (i) the environmental impact of the*proposed action."

Order Following Special Prehearing Conference dated May 24, 1978 at 5-6.

- 8 -

• 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 evaluates the Salem facility, its need'for increased spent fuel storage capacity, environmental .impacts of* the proposed modifi6~tion, ertyi~onmenta*l impact *ot" postulated accidents~

alternatives.* _ for_ spent *fuel. stor~ge, and cost-benefit balance

~f the propos*ed -modification.*. U~der the heading ii.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 will be no s1gnificant environmental impact attributpble to the proposed action other than that~which has already been predicted and described in the Commis-sion's Final Environmental Statement for the facility dated April 1973.

Therefore, the Commission has found

• that an. environmental impact state-ment need no.t be prepared, and that pursuant to 10 CFR 51.S(c), the issuance of a negative declaration ~o 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 lC at 6-12).

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 environmental

. . 13/

impact* s*tatem~nt~ - Even. were- we to assume that this action could be cons1dered a "major 11 one, *the Board agrees with the

. Staff th~t*the *evidence of record.establishes

. . that the in-cr~ental.imp~cts associated With the CC?nStruc-t;ion, .installa-tion of the racks ~nd operation of the spent fuel pool with 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. Ap:groval of the amendment will not 14/

result in the generation of any additional spent fuel.-

13. The*action by the*comm.ission in adopt:i.ng a State-ment of Interim Policy on Nuclear Power Plant Accident 10 C.F.R. §51 .. S(a).

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.

-*----------- ------~------**

• I

15/

Considerations 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 i;he cons.truction, . installation

• *and use* of the new racks require the preparation of an 16/

. enyironinental impact 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 SO 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 Ii of a new or supplemented impact statement where a final state-ment already has been published. Neither does the new policy require the preparation of an EIS where none had previously been required.* In other words, there is nothing in the new policy which suggests that. the Commission intended to transform an otherwise environmentally nonsubstantive amendment, i.e., one not requiring the preparation and The Statement was approved by the Commission for publication in the Federal Register at its meeting on May 15, 1980 and published in the Federal Register on June* 13, 1980 (45 Fed. Reg. 40101).

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 Polic;y. See n.34, infra *

.. . * \,

publication of an environmental impact statement under the criteria :,contained in 10 C.F.R. Part 51, (such as the licensing of an expande*d spent fuel pool) into a "major Federal action sig_nificantly affecting the quality of the human environ-

. ment." To* interpret the Statement of Interim Policy dif-ferently would have** the eff°ect *of* requiring the preparation of an. ._.ep,v.i.ronmental impact.* : ~tat~ent for . evecy minor amend-ment to .an .*operating :license, a* "bootstrapping" result clearly not intended.

1.4. 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 circumst-ances where 1

Class 9 risks 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-viromnental risks of accidents expressed in any* previously issued Statements, nor, absent a showing of similar special cir-

• cumstances, as a basis for opening, re-o enin, ore andin revious or proceeding.

17/ 45 Fed. Reg. at 401.03 (emphasis supplied).

As* a factual matter, there is nothing in the record of thi-s

• proceeding which establishes such "special .circumstances."

Accordingly, there is no need for this Board to consider the enYironmental consequences of what were .formerly referred

• to* as "Class 9" acciden*ts, nor is the Staff required to

.* amend or supplement its* E'.ES *issue¢! at .the operat4,ng. l~c~nse stag!3,_0r to issue~ environmental impact sta:tement in

...

~ . .. . . . . - .. . .. . ... ~.

• conj.unctio*rt.with its consideration of the instant license amendment solely to. discuss "Class 9 11 accidents. We there-fore affirm the Staff's determination to make a negative declaration pursuant to the Commission's regulations, 10 C.F.R. §§51.5 (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 19/

consider alternatives.~ Thus, it is not necessary to 18/ 10 C.F.R. 51.5 (c) (l)i prov.ides. in.*pert:i,nent part:

[I]f it is determined 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 .§51.S(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 environmental impact statement for the particular action and that an environ-mental impact appraisal setting forth the basis for that.determination ts available for public inspection.. Nega-(Ft. 19/ cont. on next page)

1-*

discuss or choose among available environmental alternatives 20/

on the basis of the record in the proceeding.- In our findings, supra., we have determined that the adverse environ-mental. impacts associated.with this license amendment will

.... 19/ (continued)*.

. tive. d,ec:J.ar;9.tions.. wili be published and

"mad~- *publ_icly available in accordance.

with § §51. 50 (d) and -51. 55 *.

• Lis*ts of.

• negative declarations will be main~ained and made publicly available in accordance with §51.54(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 action; (2) A sUlllillary description of the probable impacts of the proposed action on the environment; and (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 ** * *

• See Consumers Power Company (Midland Plant, Units l and 2), ALAB-458, 7 NRC 155, 162-63 (1978).

be* negligibly *small. The impacts of any alternative. mus.t be equal or greater, and it has been held that "[a]n alter-native which would result in similar or gre,ater harm need not be discussed." Sierra Club v. Morton, 510 F.2*d. 81.3,. 825 (5th.Cir. 1975). We therefore need not consider alternatives

. . . . . 21/.. .

or the* ne*ed *for the_ modification in* any detail. . Indeed, *

. in the . opinion. . of this Board,* not .only is such consideration

. unnecessary;* it. is inadvisabl'?, since* it i.nf:ri~ges upon prerogatives and duties of corporate management outside our statutory purview. To be sure, were there substantial adverse env*ironmental. impacts, our duties under NEPA would.

require us to balance them against benefits and examine less damaging alternatives. But where, as here, the proposed action has no such impacts, we can leave considerations such as-economic advantage and capacity requirements to those within the licensee to whom such decisions are normally 22/

. entrusted. -

21/ Id.

The Board would, at this point, note that in response to the NRC's "Notice of Intent to Prepare Generic En-vironmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel" (40 Fed. Reg. 42801, 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 contested matter subject to Board review, the Board has satisfied itself that the discussion therein meets

{Ft. 22/ cont. on next page)

Colemans' Contentions 2 and 6

2. The licensee has given inadequate consideration to the occurrence of* ac-cidental criticality due to the increased density or compaction of the spent fuel assemblies. * . Additional consideration of criticality is required due to the* follow-

• ing: *

• A... deterioration.of the neutron.absorbtion

[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 bundles1

6. The licensee has 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 considerations 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 five factors set forth in the Commission's "Notice of 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.

17. The increase in design capacity from the present 264 fuel assemblies ~o the new capacity of 1170 fuel as~'*

semblies would be achieved.by installing new racks with a decreased spacing be:t:.ween -fuel storage positions. The old : * *.,;.,

• . racks had*a nominal* center-to-center spacing between fuel

. *storag*e 16cati*ons of *21 *inches.. The .. new **racks would be

• • modular stainless steel structures with individual storag.e

. cavities* to* "provide' a nomina.l c~nter-to-center spacing _of 10.5 inches. Each stainless steel wall of the individual cavities would contain sheets of Baral, a trade name for 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 building (SER, Exhibit 6B at 1-1, 2-1; Exhibit lC at 4-6b, 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 project management and the quality assurance requir~d during the design and construc-tion stage of the racks (Tr. 602). The Licensee has audited the quality assurance program of the Exxon Nuclear Company (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

isotope per axiaL centimeter of fuel assembly, the Keff, which is a measure of the approach toward criti.cality of an array, is equal to or less than 0.95. Keff would have to be equal to or. greater than:* 1. O for criticality to occur.-*.:, .::* The

0. 95. st~dard has been se't to give sufficient margin to

avoid. crit~~ality** ~orisidering

. ..

• calc~lation~l

. . *t~chniques an.cf

..experimental veri£ication. * (Tr~ .65.7~5~; ,.Exhibit GB .. at 2-1; - .** ...

- Exhibit *le at 11-21) ** The_ accuracy *of the calctilational

• 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 Bora! 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 tiken for the effect on criticality resulting from the presence of the boron in the spent fuel pool (Tr. 550). During the refueling 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 lC 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 Keff. would be equal to or less than 0.95 without any Baral plates whatsoever if credit were taken for the Boron 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-less steel. and *the. Boral material se*a.led between an -inner and outer stainless steel shroud (Exhibit 2, ,r2; Exhibit 11, Response to Question 13}. The stainless steel shroud pro-tects the Boral from exposure to the spent fuel pool water environment (Exhibit le at 31; Exhibit 6B at 2-12). Type 304 was chosen for its compatibility with the spent fuel water, which contains boric acid at a nominal concentration of 2*000 ppm boron, and is the same material which is utilized in the present spent fuel racks (Exhibit 2,. *.,1.2; Tr. 736-38). Stainless steel of this type has been.widely utilized in the nuclear industry (Exhibit 2, ,r,r2-3; Tr. 456).

22. Based upon experience gained in environments similar to the Salem spent f.uel 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

0 at 10}. Boral material has also been exposed in water for periods up to 20 years without -any sig:rificant deterioration (Weeks at 3

• and BNL-NUREG-25582 appefrided to the testimony; Tr. 662-63). No significant deteri:oration.is expected for the b*oric acid environment in the spent fuel pool* (Weeks at

. 4) * . *':

. . . .. ~

23. Th~*.;Li.censee.

has ma,d$.. detail~d an¢i. comprehensiv~.

• .: pla.ns'. to .assure* *tha:,t 'the fabricated* ra~ks 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 crarie to assure that the installation complies with*all requirements (Tr.

597; Exhibit H [Handling, Shipping and Receiving Inspec-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 installation. These loads are significantly lower than other loadings for which the racks were designed (Tr. 478-79). As part of the quality a*ssurance 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 95% 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

at Salem to assure that all design requirements are met (In camera Tr. 26-55; Exhibit le 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 lHl *. The quaii ty .assur~c.e program *i,ncludes

• I

• I * * * '

a helium leak tes~

U:tilizirtg.

a he.limn mass spectrometer which is capable.

o:f *. det~ctin~- :very -~mall

. p°in hol~s, *. 'smaller than. any* which :.

would be significant in the fuel storage cell service en-vironment (Exhibit 2, ,r 6; Exhibit lH) .

25. In addition to and not withstanding the efforts to 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 gas (Tr. 437). The phenomenon would be self-limiting in that the aluminum oxide film formed at the same time as the hydrogen gas is generatecf is inert and slows down the cor-rosion rate (In 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

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 pre*ssure *. If a* ie~k were *hypothe.sized .to occur at the bottom half of the storage *cell,. ~he pressure would ca*use the inner shroud to

.* ..* .. bulg.e a?1,d;" move

. . toward the center*

of the cell. If the leak*

• • had :occurred. at the. top* half; no bulging would occur . (Exhibit 2, ,11; Exhi.bi t 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 fa*ilure to. be able to insert the fuel, thereby losing the affected cell from service {Exhibit lH

[Corrosion Swelling Effects];, Tr. 579, 605). Prior to loading fuel in any location, a procedure w~ll be utilized to determine whether cell, swelling exists at that location

{Exhibit 2, ,rs; 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 could 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-

sure 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 20,000psi, substantially below the

-: *...' * *. :yield. stress. -of .

55,000

• .~ ..

psi for*

. . . . uni,rradiated

~ircaloy-4

• ,. ~ubing ,(Exhibit .1H*.[Test Report. at 1-41; .Tr. 941, 743, 745-46;

• .. **Exhibit lI)*

• .'. -:Irr*adiated.

zircaloy

• wouid.

have:

even a higher yield stress (Tr. 753) ~ . Th*e 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, 119; 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 ~xamined the ability of the Baral to withstand the spent fuel .,pool env*ironment under a variety of conditions and considering a range of Boron pre~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 (In camera Tr. 10-13).

Inasmuch as the performance*of Baral is well documented in

s~gnificantly more severe radioactive environments, it was not necessary to simulate this factor as part of the test.s (In camera Tr. 13-14, 33-34; Tr. 603-04, Weeks at 4).

The coupons were examined for corrosion rate, pitting,

. bonding, edge. attack and. bulging. Two different phenomena

• wer*e * ;identified (IIi :camera *Tr~.* 21_;;23

~ . .

)**.

• One. was. a random* and

• very_ loca~.i.zed b~l9"irig

• which is.* self-:-limiting. (In camera Tr.

• .**24,*47;*Exhibit 3*and 5 at ~~3,*4~7*,*4"".'a*: Tr~ 615-16}-.

These localized bulges normally occurred in an environment different than will be experienced at Salem Unit 1 (Exhibits 3 and S' 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 (In camera Tr. 22-23; Tr. 693-94). The other phenomenon involved was the pre-viously discussed bulging of the storage cells due to oxi-dation (In camera Tr. 24). The experiments.showed that simulated storage"cells with a leak-simulating hole will sustain aluminum corrosion which will consume only a small percentage of the aluminum in the B9ral 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 (In camera Tr. 39-40).

Moreover, while some pitting, edge attack and internal gas pressurization could occur to Baral plates, B4c particles,

wnich 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 (Exhibit 2, 1110; Exhibit 8 at 1-5;. Exhibits 3 and 5 at 3-2 to 3-5; In camera Tr-.. 17,.* Tr *. 566,. 665~~6; Weeks at 2~ 4) .* As a result *of its

• • expe.r:imeri.tal : war~ . ~nd on the basis

• of previous work, no mechanism. _has been identified ~hich will d~g~ade the Baral

. .~.* .

material' "for the pu:i:pose .for which it was designed in the Sal"em facility (Tr. 435-36).

29. The Licensee, in addition to these test programs, has committed to a long term fuel storage cell surveillance program to verify that the spent fuel storage cell retains 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 Surveillance Program]). Samples of -flat plate sandwich coupons and short fuel storage cells will b~ provided for periodic surveillcm.ce and, testing. The samples are of the same materials and are produced using the same manufacturing and quality assurance procedures specified for the fuel stor~ge cells. One short fuel storage cell and one flat plate sandwich coupon will be prepared such that the Baral 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

years thereafter (Exhibit 2, ,111; 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. 'l'h,!=r.e:. wa's consider.ab.le. discussion ~uring. the

• • • course.

of: the.'hearing_ . ..

concerning

.problems

. encountered at.

: otiler op~rati~g . facilities . regarding .these . spent . fu~l r'acks.

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, nondestructi_ve testing, and long-tel:l11 surveillance programs for the fuel pool, assured that problems which have occurred at other facilities are not likely-to occur at the Salem Generating Station (Exhibit 2, ~12; 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 Baral. The periodic sampling and testing of the Baral coupons would detect any incipient deterioration. Thus there is no substance to the Colemans' assertions regarding Borai.

31. In. reaching : this conclusion*,* the Board has con-*

• . side;red the r~levant. *portions:

. .... of a report dated, April 10, 1979, made by: the* Commission's .Of_fice of Inspection and

-E~forcement

-o:0: it~ _*fipdings* *relating. to an inspection *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 bee~ caused by swelling of the cEall walls due to the bu*ildup of gas re-leased within the walls by a chemical reaction between water and the Baral material. After the cells had been removed from the pool,_vented (by drilling holes 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 dununy fuel element, however.

32. The Monticello racks were not designed or manufac-tured by Exxon Nuclear, the supplier of the Salem racks and were of a different design (Tr. 458-59). Each of the af-fected storage locations was resized "by vacuum and mechani:cal means~" a procedure which has not been proposed for Salem

• As. 'previously :discussed* at paragraph 26, at Salem, were an empty cell to :swell, it.. would be *considered unavailable for .

use (Tr. 580, 605, 609). Ther*e are no plans to return such a cell to service. If a fuel cell at Salem which had a fuel assembly stored in it developed a leak, a completely dif-ferent situation would arise. With a fuel cell stored inside, a fuel storage cell would not bulge beyond its elastic limit and would thus return to its original shape when pressure was removed by venting (Tr. 606).

33. Moreover, there are substantial design differences between the spent fuel racks at Salem 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 dmnmy fuel assembly and presumably an actual fuel element, there has

been no showing that this has any significance to the present J;)roceeding. The Colemans have fa'iled 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 .*

Sa.;Lem. .

?S. , D~ring __.tl?,~ course of .the proceeding, intervenors 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 determination is based upon the assurance of leak tightness previoµsly discussed and upon the ability to vent the racks

• at some later point in time should the rieed arise.

LACT Contention 1 The Licensee ,has not considered in su£ficient detail possible 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-rangem~nt with Allied Chemical Corp.

be stored at the AGNS Plant in Barnwell, South Carolina.. Furthermore, the Licensee has not explored nor exhausted the p~ssibilities 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 *the record, the Board has addressed the merits of this *con:tentlo~* should a reviewing tribunal disagree with

• thi*s Board,*~- analysis~

• As an initial matter, the Board vie~s the second s~nten*ce of the contention a:s limiting the

• first, applying the principle of statutory construction, expressio unius est exclusio 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 f.irst time.

• For the reasons discussed below, the Board finds that alternatives to the proposed expansion of the capacity of the Unit l spent fuel pool have.been adequately considered by the~Licensee and the NRC Staff and that none is superior to the one silected*by the Licensee.

37. The question of when the pres~nt 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

re£ueling outage occurred 22 months after fuel loading at which time forty, instaad 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 *. 110.4-*06) ~

.The'p-lant spent fuel pool would thus be completely*fuJ.l

-after the fourth refueling which is scheduled in 1983 (Tr *.

162~, 1030; 1104~6J. Wit~ *regard to the discharg~ of 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 dat~*and which may reasonably be expected to occur in the next few years would not significantly affect the Board's reasoning concerning its review of alternatives (Tr. 1029-40, 1043-45). In this regaz-d, the Board cannot ..ignore the 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 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

. *-: ... ;:\~*iven***n~: e;,id~h.~i°a:r.~/ *weight*:~~. ~e ..resp*onse to* these; qU:es~:i.oD:s*~

• • . See** Pacific Gas:

and Electric Company (Diablo. Canyon Nucl.ea~

Power P.lant, Units Nos ... l and. 2), ALAB-334,* 3 NRC 809, 825 (1976).

39. It is not practicable to store the spent fuel from Salem Unit l a t 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 tress-examination that the 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 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.

li*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

. approa~h. In ;any_.event, using an expanded pool at Salem Unit 2 in conjunction with the existing racks at Unit l

.-_ . :.. :_.. .. : would: _only.~ permi*t storage until 19 9-1 .(Tr~ 10 3 9-4 0}

• 4 o*-~

• Moreover, . 1:he envir_onm.e*ntal impacts . of *the extra

• handling. of. irradiated spent fuel., such as the dos_e received by workers during the transfer, would have to be attributed to this alternative inasmuch as the spent fuel pools for the units are completely separated and each fuel element would have to be removed from the Unit l 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 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 COlJiltervailing advantages to this alternative. None has been demonstrated to us. Moreover, if the ~nit 2 racks had to be changed to the greater capacity racks to increase storage at a later time and Unit 2 fuel had to be shipped to Unit l for storage, the doses would be even greater in comparison (Tr. 1138-1152}. Due to the uncertainty in the availability of an Independent Spent Fuel Storage Installation ("ISFSI") by that time (Exhibit 6C; Tr. 1005), such an alternative could impact adversely on Unit 2 operation; the Board considers it

- 33 -

as* only a short-term temporary expedient and not a substitute" for.the proposed action.

41. With regard to storage of Salem Unit l spent fuel at the,:Hope *Creek units, i.t is unlikely that these units would be sufficiently complete to enable fuel to be stored pr:i~*r *to** tli~-:urim6d+/-f_i~d-* Sal.erri *1iriit b*~irig. full** (E~ib-~t '6C at:'

~~ ~ . *~~~/ ..... --~~~~a~.e ~~ iiop~ ...Cr~~~

1*

• f 1*;*~1~**;

~ould involve repl~cement of the Hope.Cxeek racks with racks- capable of holding Salem l fuel, further limiting storage capacity for spent fuel generated at those 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 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 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 rio adverse environmental impacts 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 stated that in no event will the facility be utilized by

- 34 -

AGNS for the storage of reactor fuel absent reprocessing

. (Exhibit 2, ,121; Exhibit 6C at 14-15). Considering the President's April 7, 1977 statement deferring indefinitely commercial _reprocessing and recycling of the plutonium produced iri. the u*. s ~ nuclear . power programs' the storage

,, -_:*: '>.*~~l?~~i.~¥-. o:f. 'tha.t :*:1:~c.ii:1:ty*. *c;:~Iin?t*'..-be-<r:eJ).ep.:_~pon_;*. ' ..... : ... ....... .

.**. 4-4~ The .N~C ha,d. unde.r review an application. by Exxon Nuclear*Co. for a'.s'torage*pooi' 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 l is not prudent (Exhibit 2, 1122).

45. The fuel storage pool at the Morris, Illinois 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-4 process. Other spent fuel is not being stored in the absence of an express commitment to do so. There is no such commitment for Salem (Exhibit 6C at 14; Exhibit 2 a.t 1123).

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 1123) *

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 authorizing construction of such a tacility (Tr. 838}; neither has the environmental impact study of

• :***su~h. *a. project been** q9mpleted (Tr. 838 ~- 980-82).. Further~*

. mo~e, . n*o site. has*.. been

• .s~lecteq

{T.r. 844)

.. . Recently, . tjle Department of Energy indicated that th1:re 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~ 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 l inasmuch as a new pool structure with special design requirements, i.e., some seismic resistance, and

- 36 -

sup_porting systems would have to be designed and constructed, and S?ent fuel transported to such a facility (Tr. 986; Exhibit 2 at ,124). Even the witness for LACT has recognized this (Testimony of George*Luchak, Ph.D., Transcript follow-ing 918 at *2-4 [hereinafter "Luchak . at - 11

]).

• In addition,

.**i *: ."*latid *would. have

• .t6 be. ac;quired i *.and .*it*. w~uld* be*:. nec~~s~ry.:

to ..

over':ome. li~ensing .pr.oblE=l1ls . ~E~ibit 2 .~t ,t 24) *. The environ-me~tal _impacts -*~_s.soqiat_ed* with con~tru~t-ing *such* a. :cacility would also be greater than the minor impacts associated with replacing the racks (Tr. 790, 835, 977-80, 1083-84). The proponent of this contention,. LACT, presented the testimony 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 engineering textbook (Tr. 895-96), has not written any scientific article~ concerning this subject matter *{Tr. 896-97), has never visited the Salem.facility or any other nuclear power plant for that matter (Tr. 897), has never* studied the layout of the plant and was not acquainted with the detailed design of the plant or fuel pool {Tr.

900}. Furthermore, the Board has found Dr. Luchak to be totally unqualified to perform any accident or radiological

assessment 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." *852,* *954*-ss*,** 987) .* *.* 'l'he Board has determined that

• .. -..... ~. .. ~ . . . ... : . . *. _.. . . . .: :. ... .. *; . . . .

. .. . . . .~ . . .

in_so.~a:r as. cert~~ ,.acciden.ts .were nqt includable in the

• . *.. design basis -by. the. Gommiss*ion at the. operating license stage, their postulated impacts need not be considered as part of the review of alternatives (Tr. 1047-50). In any event, Dr. Luchak is unqualified to provide any expert opinion regarding accident probabilities, ~onsequences or risks and has given his testimony no weight. The Board finds that an ISFSI, particularly one in a distant desert site, not to be a superior alternative.

48. All previously discussed alternatives assumed that the spent fue~ pool could be filled prior tG the alternative being needed. This is no*t quite the case. After the next

( s*econd) 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.

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 nuclear power, it is unlikely that permission would be . ::'~:

granted*to export spent nuclear fuel.* In fact, the President's*

, * * = _: -* .' .**-: :ipr~1-*7:; .1971 s~at:~ent on ~uci1ear* p6~er policy_* stat~s* th~t -*

the U. S ~ i_s explor'i_ng .."meas:ux:es, ~0- as*s~e acce~s to* nuclE;ar

_- fuel '-~*uppi~-e-s -~~d- spe~t- fu~i- st~rage [in .the *United States]

for nations sharing conunon non-proliferation objectives."

More recently, as part of his program on radioactive waste management, of which this Board has taken official note, the President has stated that the U.S. would accept "limited amounts of foreign spent fuel when the objectives of the U.S. nonproliferation policy would be furthered. 11 Thus, the Board finds that disposal of Salem spent fuel outside the United States is not a viable alternative.

SO. Inasmuch as the Commission has at j:he 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" 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 should a reviewing body disagree, we have included a dis-cussion of this matter. The Licensee has estimated that a

shutdown of.Salem Unit l with a net electrical..output of 1090 megawatts would cause incremental repTacement power costs alone of $500,000 per day, based on the differential costs.of producing energy.from Salem as . compared* to.produc-ti~n from other available* units in the PSE&G an¢iPe~sylvania*

~e~:-Je;~-~-;: .~a~;i.~~~- **c**.~PJ~i,_.) r~te*~~:orui~~~~o~- (E~ibi~--2. a~::* ... *.

-- . 1127). , .The ~.taff~ .looking. at the, long term. economic. impacts

.. other than 'the. short t~rin i~cremental effects' . facto;ed in .a.*.

capacity factor range of 60-70% to arrive at annual replace-ment costs associated with the discontinuance of operation on the order of $300,000 to $350,000 per day (Exhibit 6C a.t 18-19). Using either figure, these costs would still be far in excess of the costs associated with the proposed modifica-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 i'f the application to expand the

~

spent fuel pool were granted 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 to leave such a determination to the Licensee, based upon

the info:i::mation which is ..available to it at the time a decision in t..'1.is regard is to be m.ade, such as the availa-bility of pe:i::manent disposal facilities .

. 52. In conclusion, should the Board be required to review alternatives to* the proposed action it finds that

• .**.**.*other *~v~i°l~le**~1te~nat.ives *are not p*referable to the

• ins'.tallation of the_ in~~eased capa_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 l 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 "meltdown" at Three Mile Island, 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

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 ..secprid ques::tion. as.. unnecessa_ry to thi,~ proceeding and

~-~ \:im~-* wh:e~-*-;e :~o~i~:~-he~:.: ~~ide~~e* .. ~~- *.th.~ *.o=~~;

po~~pb~~*d.

questions.. The. Staff°, Joined by. the Lic~~se~, * "aiso .. obj e~_ted

. 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 identifx 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

spent fuel racks has been identified and the existing design 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 accident was a Class 9- accident:

• . *..... - :. ... ,. :_.* ':['.he* .prop.os*ea .:An~*ex*~ to AP~-~ndix o*, 10.- .

.. i * .

. *. CFR 'Part** *.so-; .appears to define a Class 9 accident as*a sequence qf _failures

_' *. w~ic~ *a~e=- m9re severe t'.han. those which.

• . the safety* features.* .of* ~e plant are designed ~o 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~ Class 9 accident? Was the risk to health and safety and the en-vironment "remote in probability," or "extremely low 11 at Three Mile Island, as those terms are used in the Annex?

{Tr. 922-23).

57. On February 22, 1980, after receiving the parties' varying responses to its fourth question, including the Staff 1 s answer characterizing TMI as a Class 9 accident, a position which was not free from dissent from within, the Board issued its Februart 22, 1980* Memorandum and Order ad-dressi~g, 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),

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 fifth question to the parties for.consideration at an evidentiary session:

• In i;he

• event of a. gross 1*oss of water from the storage pool, what would be the differ~nce in consequences petween

..; .those: occasioned by the pool. with ex-

. P.anded .storage* and those occasioned

• by *the *present pool. (11 NRC at 346)?

59. In our Memorandum and Order, 11 NRC at 345, we made 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 Licensing Board*' s fifth question was proper. This request was denied as premature in Public Service Electric and Gas Company (Salem Nuclear Generating Station, tJnit 1), ALAB-588, 11 NRG __ , (April 1; 1980). 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 session which permitted this Board to directly apply the generalized treatment of Class 9 accidents to the factual context of this proceeding, a further explanation and analysis of the Board's position is appropriate.

60. On March 21, 1980, the Commission issued its Memo-24/

randum and Order in the Black Fox proceeding;- in finding

• 25/ -

that the Appeal Board- had misinterpreted its decision in 26/

Offshore Power~- the Commission vacated that portion of ALAB*-*573' r;.;hich directed 'the* Staff 'to f~ie it~ views on

. *.wheth~'r ..cia~~ :9 .* ~~-~~,d~n~s *.:.:

~h~~ld' b*~ co~~idered in the

.. 27/

~lack Fox _proce~~ing.-.-

. 61 *. Th*e Commission. reiterated tha't' 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 developed by rulings on a case-by-case basis because "piecemeal considera-28/

tion is not*appropriate to such an important policy area. 11 -

It added:

Because the*existing policy on Class 9 accidents was not displaced in Off~

shore Power and would not be dispiaced 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 (March 28, 1980), which gave the Licensing Board instructions consistent with

• the Commission's decision.

Public Service Company of Oklahoma (Black Fox Station, Units 1 and 2), ALAB-573, 10 NRC 775 (December 7, 1979).

Offshore Power Systems (Floating Nuclear Power Plants),

CLI-79-9, 10 NRC 257 (September 14, 1979).

CLI-80-8 at 1-2.

Id. at 3.

pending generic consideration of Class 9 accident situations in- policy develop-ment and rulemaking, the Commission en-'.". .,

v-isioned that the *staff would bring an*".

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/ .*

.. 62. T?ius,. t.he Co~i~~ion *ha~ ;e~erv~~ to 'its*eif. the dec*ision of whether Class 9 accidents may be considered in

. 30/. .

any* given case'.-

• Absent afffimative action *by the Coni~-

mission, the Staff and licensing boards are prohibited from proceeding with any such consideration.

• 63. The ConnnissionJs Black Fox decision also specifies that the recommendation to the Commission must come from the Staff, which has an affinnative duty to bring to the Com-mission those cases which it believes warrant consideration 31/

of Class s*accidents.- We interpret the Commission's Black Fox decision to mean that the-Appeal Board, and by implication, licensing boards, are not empowered to order the Staff to make,such an affinnative recommendation to the 32/

Commission.-

• The Appeal. Board found this "unaml::liguous ex-pression of Commission policy" controlling in this Salem

. 33/

proceeding.-

29/ Id. at 3-4 (emphasis suppiied). See also id. n *."3.

30/ ALAB-588, 11 NRC a t _ (slip op. at 9).

31/ Id.

32/ Id. a t _ (slip op. at 9-10).

33/ Id.

64. As part of its response to Question 5, the*Staff presented its views with regard to whether,. in a:ccordance with the criteria set by the Commission in Black Fox, 34/

supra, Class 9 .accidents- should be considered in.this 35/

proceeding.- It st~ted that "[t]he expansion of the spent

• **fuel pool at the.*salem* site 'does not constitute an exceptional case

. .*.** resulting ip risks .substantially

, greater than for an average.plant" and its ultimate conclusion that "the environmental consequences of Class 9 accidents *need not be 36/

evaluated."-

65. In its February 22, 1980 Memorandum and Order, the Board drew aµ 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 relea.ses were a Class 9 accident to be postulated. As a The Board and Staff witnesses have utilized a defini-tion of Class 9 accidents consistent ~ith that noted in ALAB-588,'n.2, ll NRC at (slip op. at 3). See Tr. 1464 and the Board's Memorandum and order, LBP.::-

80-10, 11 NRC at 342-43.

The Board recognizes that the Staff's silence wou.ld have been sufficient to pre.elude consideration of Class 9 accidents and that under ALAB-587, Black Fox, the Board could not have ordered the Staff to present its views.

Direct Testimony of Walter F. Pasedag in Response to Board Question No. 5, following Tr. 1387 *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. §2.206 which at pp. 28-39 addresses in detail why the Staff does not believe i t appropriate to con-sider Class 9 accidents for this facility

• result of the evidence before it, the Board con*cludes that it is clearly not appropriate tq .* consider Salem as having a liquid pathway comparable to a floating nuclear plant; rather such liquid pathway i 9 c~mparable to that of'the land-based plant used as a bas.is for comparison in the '

Liq{i:id' P.athway

,*c;en:~ric

~tudy

.(NUREG-04.40-) **

. . . . G~o~d~ater transport,* sur.fac*e water transpo~~' and usage of ~e- w~ter .

. . bodies* -~-u~roti~di:ng. t~e- _sai~- ~ite* w~~e exam:i.n~d (Paseda~ .' 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 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 more conservative results .(Tr. 1600-04)

• Thus, the Board is able to conclude that special consideration of Class 9 accidents at the; Salem site is not necessary because Qf any unique

. 37/

considerations associated with the liquid pathway.~

~6. The Commission's Statement 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 the liquid pathway, none of the other parties or participants addressed this as a signifi-cant pathway resulting from the "gross *1oss of watern in their preferred testimony.

- 48 -

July 1, 1980 date after which additional discussion may be needed in an impact statement not completed by that date.

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,

... ,.* *.. th~** st~teinen'~

-~iio-~1d ~bt ..se~e

.. a~ a :basis** for ~pening ;* * .

. . . .. . reop~ning.

or*..e,xpanding any previous oi. ongoing. proceeding

.* . ~o.reover ,** the Cortunissidn *.express*~a. c~nfidence* *that; even.

i£ 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 Boardis next consideration was whether the "gross loss of water" postulated by Question 5 was an ac-cident that could be classified as orie 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 that 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-tect-ion arises from the pool modification (Pasedag at 2~ Tr.

1463-64).

38/ See p. 11, supra.

39/ Statement of Interim Policy,. 45 Fed. Reg. at 40103.

- 49 -

68. The Staff testimony analyzed the weld channel leak detection system: in terms of an initiating mechanism for a "gross loss of* water.'.' The Staff explored a scenario which 40/

involved multiple pu:;-ict1:1,res of the fuel pool liner-

.: ieai:ling. to a. maximum* leak rate of no *more than *110 .GPM cal-.-

culated on' k' co~se;vative* basis . . a~d-*p*robabiy

. subs.tantialiy*

. less, which would be.detected and ~ppr9Eriate ~~tion taken

. . *. . .* . .:  : .. * . . . *. *.' . : . . . .. .. . = .* .

,.. tcr preci°tide* uncovering of any of the. fue*l* (Pasedag .-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 causation of a "gross loss of water,"*we beli~ve it profit-able to review the initiating causes as advanced in the testimony responsive to question 5 to set tile consequences in perspective and to view it in terms of other risks as-The Licensee's testimony, Exhibit 14 for identification, was similar in this regard. See also Tr. 1470.

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.

sociated with the facility. As*explained previously, the Board has** no't felt it necessary to re-review the design basis of the facility for its protection against natural i,

• I I

phenomena~*.,_. ,i.rone of the parties have advanced any. reason for.

• .doing so. Dr. Webb, a witness for Lower Alloways Creek

• Township, has advanced the idea of an earthquake larger than the. N~.9 h,as requ:i:r.ed

. . . . ;qr, the de.sign ba_sis as causing .*.n a.

gross loss* of water. II The. Board 'believes that such a hy-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 42/

NRC's requirements contained in 10 C.F.R. §73.55.-

Neither does it believe in view of the findings it has made with regard to criticality and Baral, that it need consider an initiating event deriving tram 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 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.

• 42/ Commonwealth Edison Company (Zion Station, Units 1 and 2), LBP-80-7, 11 NRC 245, 283-85 (February 14, 1980).

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 discussed, -infra.

• With regard to the present discussion, suffice it to-~ay that such accident would not have had a

• _. * ..sig-nifican~- _e~fect_ on. the spe~i:. fuel pool; it would -take an

• • accident* sign:i-ficantly 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 facility or to serve as the basis for disclosure of environmental impacts.

71. Having given further review to the responses of the pa~ties and participants ;to the Board's fourth question 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 with 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 a quantitative *value on the probability of an accident

'.* *serious eriough. to: affect* the* spe~t fu.el p*o~l, the Board

..bel~eve~* tjlat. sue:h c;Ul_* .a~cident.. i~ extr~mely improbab:ie when ..

• • • • * * -.'.: . . ~i~~we*d :ih _. ~h~ *.: ~~-~:ct~~\; f. * ~c2i~e~~s *th~\: *h~v~ b~e*ri pre*vi~~-s ly placed in the 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. 11 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 11 to be extremely low. Specifically, the consequences of such a 11 gross loss of water 11 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 11 7asedag Further Testimony at_").

This combined with the predicted low probability of occur-43/

rence of the _"gross loss of water"- permits the conclusion As discussed above, *the* Board has not identified any specific sequence which could cause an instantaneous

. "gross loss of water. 11 Moreover because of the ad-ditional series of low probability events necessary (Ft. 43/ cont. on next page)

th~t the risk associated with such occurrences would not, in the Board's view, constitute a significant environmental risk.

72. In_s~ary, this Boa.rd believes that the initiating even.ts which coui"d cause a gross loss of water are cumula-

."tJ.vei.y ~o. remote as to .reduce the risk to such an extent as

. * -~9. req-qiz:e no _-fui;-the_r discus;:;ion.

• How~ver, *in an abundance

.. *:..'*.:.:- .* o;* -~:~ut~ori;** .th*~;:~o~;~ :*a~*~u~*s*~~-:;th~**_evicie~c~---~eio~e .*it 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-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-

~

saning 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 o*f record, including testimony of the witnesses and giving appropriate weight to each of them, as discussed below, the Board 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.

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. *s.;**

• sto.rage* of *fuel in the present racks is not substantially.

• *. * .... * .* 44/. * ... , *.*

greater.~ .Th~ range of consequences and.probab~l~ty has.

been. developed. .

in the.-*

. .. record of this proc~eding. *.

• .* * * *-*t4~: ***.The "iead* ~ifu*~-~~**-~*a'~ Walt'~~- --~~~edag~-

• The Board ~~s**

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 44/ The risk of a gross***1oss

• of water has been previously discussed; 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 paragraph 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)

55*-

apprised of the analytic techniques utilized in the heatup analysis. The Board found his testimony generally to 1?e cogent and well thought out. Assuming that the most recent-:1.y discharged rods had reached the temperature for self-sustaining oxidation, *or. Benjamin wa*s unable to.conclude whether suff°icient heat 'would be . t;~nsfer'red to *spent* fue1** rods whi~li ha*a. b~em _stored four years and longer to permit these

.-;~d~--::~q.: r~~ch-~ th.e . s~lf*~*s*~~~*a:i~ing.'. ~~id.:a~io*~-- *~~;pe~~t~r~*~.a" .

This step was not part of the Sandia Report which only examined the status of the most recently discharged batch of fuel (Tr. 1433, 1441). Dr. Benjamin stated that he was 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-corporated in all facilities to minimize the likelihood of a loss of pool water, including (1) the conservative design philosophy of building the concrete struc-ture, r1;cks, cooling system, and* support structures to withstand the forces that might result f~om a large earthquake or tornado; (2) design 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].

46/

analytic work he had done to date,~ but in his opinion, further investigation was warranted (Tr. 1436-.39) ~

• While from the standpoint oi understanding the basic phy~dcal processes involved, such an analysis might be interesting on a generic basis~ *-~e BOa.t:d conc.ludes; as* discussed :below,

• th~t ~e f~rther* a~~l;si~ * ~a~h-~t be*_ .j*~~~:Lfie~,. i~ **light of

• : . *- the evidence

. *which* has . been . :recei.ved

. in this

.

• do~ket

.

• and . ...

-*:: *-: * *:. *. *. -~orisi"id~eti~g_:*the *:l1.nc1+/-ii~s* . 81~"t

_thi~-

• B~-~d****is,.req~t;~l to*.--make* * *-*.*

• under the Notice of Hearing and Commission rules and precedents.

75. In contrast with.the reliance on. the testimony of these two Staff wi tness*es, 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 addi~ion, his testimony had negligible 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).

- .JI -

probative value with regard to the Board's fifth question.

Dr. Webb characterized many physical phenomena as "con- ' . ;

• cei vable 1

' 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* d:id ,not attempt to

.: .** attach any *prdbability _'to .each s.ilch**ev~nt~. nor:*was he*. ~le

.. to give any s.u~h estimate. from the st~nd (Tr*. 17_10-11) *.

. :* -' . *.Thu~,.. it. i~:* ~ct..;~ss:iiie: ~o* gle'an: from-:: Dr..-w~bbi*~ .t.estimo~~ .. *_*_.-_.

• whether a* postulated occurrence had* a probability of O.l, 10- 7 or 10- 1 4 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 b_e 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 numb.er of broad areas, the Board is convinced that Dr. Webb does not have the qualifications of an expert in these fields and the Board has discounted his testimony heavily (Tr. 1701, 1741-43) ~ Dr. Webb based certain of his

co..nclusions on "experiments" he did with material removed from flash bulbs; the Board does not consider this to be.a true scientific experiment (Tr. 17*54) .* Finally, Dr. Webb was admittedly un~ble to reach any:. conclusion regarding the.

• ability -~f the' present ~~cks to**:"sustain*: th*e fuel without

• • .rea:ch1~g'. ~. ~elf.:.:,_.Su~t~ining oxid~tiC?n temperature and thus to

.

• testi~y *r~g~rding *the .. di!.f.ei:'!=n~e .in c~ns~quences (Tr~* i716;

.:. *._. *.**: ., . . .:. . *:,.~e~~; *~~~~. Ii~:.*6:l :*~-~~plea~~~ :a~: *'.~*~j:.: ~~~-: ~~eri. *-R~6k\in;{i~-~-st 6£ .

Fuel Heatup, *. attac.hed 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. 1575-7~). 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-staliation and *use of the higher capacity racks, nor has

59 any scenario postulated for the racks been shown to have an increased probability of -o.ccurrence. No design deficiency nor need for improvement-has been identified (Tr. 1604-10).

77. . For fuel* freshly, discharged from the reactor, the assumed* continued* denial of*water cooling capability may

• ev~ntually lead to oxidation and failure of the clad and to ov~rh~ati~g.

of_

U02 .. fµel~. ,with

• the pot.~n:t~al for the release.

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, Mr. 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 .

the design of the*buildi~g, there would orily 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

of a c*ore is involved and there has been some decay time 47/

since operation of*the core at full power (Tr. 1447-48) . -

Looking at the ol"-ier fuel, while there is a possibility of some fission product release from older fuel, it would only be a few 1.sotopesind would be small (Tr. 1448-50, 1452, 1526-27, lEfo0-01).* *

• The Board* recognizes that the comparison

..to. a -reacto.i;- _acc;:id,ent. is .n_~t e:xact*, _but<_stil_l considers that

• • • i't .db~s* have. ~ub-~~a~ti~l ut+/-"lfty~* :*:it'\s.eriionstrates. that 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 ox*idation as a conservative criterion for the release of the fission products of the fuel in order to estimate the differences in the potential consequences of ..

this hypothetical* event arising from the p*ool modification 48/ .

(Pasedag at 4).- The new storage configuration may result 47/ Furthermore, 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 l 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 cer~ain other plants having expanded racks (Tr. 1590-91).

in*less natural convection and hence a higher likelihood of reacl'iing oxidation ternpera tures and possible clad melting *

. 4S/

for recently_ discharged fuel.- Although heating of fuel assemblies.,stored adjacent to the most recently discp.arged assemblies wouid occur, it is- the view of* the Staff, which is being adopted by the* Board, that there has b*een no credible mechanis~ . !or:. the prop_~gati9n _of a "zirconium firel'. 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. Benjamin 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 J:?atch would not reach a self-sustc!ining o.xidation temperature after 60 days

• of discharge (Tr. 1454, ~572-73). This array could be maintained until half the capacity of the pool were utilized without any shuffli~g 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.

cqnfiguration 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 niost recently disch.arged*assemblies would occur, it is the vi~*w:*o':i:' the *staff.,. which :is b*ei~</ adopt.ed by. the Board~ *-tjiat*

there has.been no c~edible. .mechanism.

for th~

prop~gation"of

• .... a: *zi~c~ni~::fire":

• or. t,he. sp;ea:dj;~g-*of.

o"ther 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. Benjamin 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-sustq.ining 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 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.

were fuel in the old racks assumed not to melt, immediately 50/

afte~ discharge should a "gross loss 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-susta_ining oxidation temperature (Tr. 1457, 1497; 1503"~05, 1508-09)

• If the newest elements*

do* . not rea~h .the. *se*lf-sustaining . . clad* oxidation tempe-rature,

. . . *. . . ~ ...*~. . *. . : **. . ....

• • .then the **_older elements also wiil 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 Pasedag that, while small, the risks associated with the movement of fuel would be higher than the risk in not keeping the checkerboard configuration described abGve (Tr. 1585, .

1587). Thus, the ..Board does not believe that there is even any reason for prescribing 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).

.I

.~

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 ass~ed .by* i:;>r. *webb iri his testimony was* that *of a "zirconium

• fire, 11

<3. .defl?l,g-ration. with columns of. visible flames .. shoot-

.ing up from* "tne spent fuel pool with. l~rge con~ecti~ri forces~

air currents and drafts (Webb's Addendum at 2-3; Webb's Part I at 31-35). When challenged, Dr. Webb retreated from his original definition of 11 zirconium fire 11 (Tr. 1752;..;53) but did not otherwise define a specific mechanism for the spread c{f the self-sustaining oxidation. To achieve the large frac-tional or total releases predicted by Dr. Webb would, as a minimum, require a physioal mechanism such as the* 11 zirconium fire 11 described above. To a large extent, Dr. Webb based his assumption of a possible deflagration o~ a fire which occurred on* April *2a, 195.5 at the Bettis Atomic Power Labora-tory in contiguous open ains 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) *

80. The pyrophoric prope~ties of zirconium .+/-n *a finely divided state are well known. The witnesses for the Staff testified that after Dr. -Webb had raised the spectre of a

~*.' .. '.

"zirconium fire" they had investigated the literature -and spoken with recognized experts in the field. Both concluded th~t -th~ Bettis fire* experience was not. applicable to the

. ~it~ation.at ha~d-because o~* the difference in -the. state of the zirconium. (Tr. 1518*, 1s21-22,* 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 demonstra_te 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.

1517). Dr. Webb extended the definition of aggregates beyond that contemplated by Dr. Baker to include fuel as-semblies (Tr. 1744-45, 1749-so*, 1788-89, 1792-93, 1799-1800). It is clear to the Board from the context of the

paper that this was an improper application and entirely unwarranted. To find that Dr. Baker's article suppo'.rted his hypothesis, Dr. Webb apparently deliberately ignored other.

portions which were counter to his position. For example:*, .

he _ignored Dr. Baker's statement that ~ingle pieces ~f**

. zirconium *heated.slowly do not ignite even*if heated to

• 1300 ° C because of. the tough ~xide fi.lm that q.eveloped*. (Tr.

• * :*. * * * >{:7~4_::3~i: .3i7.~0.)*/*:* *D*~*~. Webb al.so*::igilo~~d*. in his*. c~lculations 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 sou.ght 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 unwarra~ted; even assuming the pressure increase. took place at all, it would undoubtedly occur more slowly, causing leakage from 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., plateo~t and retention within the fuel and Fuel Handling Building.

83. It is therefore our ultimate conclusion that, 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 pool, - the *subj.ect- ~f po~tions_ of Boa,;-d questions 1* and 3 *.

. . .. : .... . The Th.tee Mile **Island Accident -and. Its. Effect on-the' Salem:Un~t. l Spent' Fuel Pool _Wei::e Tha-~* Accident*

.... Postulated to *occur at the Salem Unit 1 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 inasmuc*h as the unit was in its first cy*cle ( "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 si!ored 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 fuel pool if*the TMI sequence of events were to have occurred at Salem. Based upon published data (Staff TMI Testimony at 3), the following sequence of events was postulated to have occurred. After the reactor scram at TMI-2, which was

),

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 pressur~ transient and failed to reseat, resuiting'

. . in.. an overflow of reactor* .coolant system water

.fr9m .t~e. reactor* coolant. ~airi . t-a~

• t~ the*. rea~to_r building.

'.(conta*inmerttf-* sumi:;i *. . The r*eactor* building sump pumps started automatically due to the rising water level and discharged water into tanks located in the auxiliary building. These tanks became full and overflowed into the auxiliary building.

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).

Based on the information on the* 1I'MI 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 uranium dioxide fuel pellets and the zircaloy>cladding) in the reactor core due to clad failure. In addition; radioactive noble gases in the fuel* pellets were apparently released from the cqre resulting in a total estimated release of approximately 30 percent of the core radioactive noble gases. The activity was released to the primary reactor cooling water, a portion of which was spilled onto the reactor containment floor.

Some of.the. gaseous activity in the released reactor coolant,

particularly noble gases, became airborne in the containment structure. As noted above, ;:.some of. the spilled reactor coolant-was pumped into the auxidiary building liquid waste storage tanks, and a portion of this water spilled onto the auxiliary building floor, resulting in the radiation levels

• 1:n .the auxiJ,.iary building and substantially all of the

• radia-ti*o:p 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 from the Salem Unit 1 reactor core into the reactor coolant system. It was furthei;- 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 thi5 Board's purpose is not to determine whet~er 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

s~ages 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 about 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* tr*an:sfer. of the. contaminat~d water. in the containment sump would occur at Salem Unit 1 as it did occur at TMI-2 (Staff TMI Testimony at 3-4; Licensee's TMI Testimony at 2-3; Tr. 1268, 1274-75).

88. While, because of these and other differences.in design between 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, tjley were assumed to occur (Tr. 1185). Moreover, the physical design of the various components associated with the spent fuel pool and their locations are different. For example, at Salem the spent fuel pool is locat~d in a separate fuel handling 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

containment 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 wa-lls 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

• 1;0 be .less than 400. mrem/hour~* .- Even this radiation level,

.*. , a***  : .: ....*

which -~ould 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 w~ich 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 c9oling system components,* although protective measure*s 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 TMI Testimony at 5; Tr. 1181, 1333-34). The Salem Unit 1 auxiliary building ventilation system is designed to prevent the movement of airborne radioactivity from one

'**~ ... .....

I

• potentially contaminated area of the building to another.

Thus, any.gaseous activity released from spills in the

. liquid r~dwaste systa~ would not contaminate areas associated with spent fuel pool cooling _system components (Licensee's

• TMI Testimony at 3; Tr. 1292, 1278-79).

9 0 ** ' *. The -fans an~** *f11 tering *.equipment : ii:L .. the Fuel.

Handling. Bui+di_ng ventila.tio_D: system a;-e_ l~cated in t:he

• penetrati~n a:r~a at el~Vation lQO ft .. (i.e. I ground' level);:*

between the containment.and auxiliary building. Since this equipment is operated remotely from the Control Room, access to this equipment is nonnally not required. In any event, maximum radiation levels, which are calculated to be approxi-mately 12 rem/hour would not be high enough to preclude access. If necessary, dose levels could be significantly reduced by placing temporary shielding in front of the personnel access hatch in the *penetration area, the primary contributor to radiation level in the area. A single row of lead bricks would*reduce the radiation level to less than 10 mrem/hour (Licensee's TMI, Testimony at* 4; Tr. 1288-89).

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 §2.2; Exhibit 6C at §5.3).

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.

Moreover, the proposed activity under review by this Board*

would not significantly affect the analysis. Thus, regard-less of the absolute ~ignificance of the postulated event at Salem, it is insensitive to the amount of fuel stored at Salem. This fact alone would satisfy the Board's limited

. inquiry (License.e's TMI *Testimony at 4; Staff TMI Testimony

_c!.t 6) * '* * . * . !';,. *' ,* ...... .

. *. 9 2. ***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, i t would be negligible

{Licensee's TMI Testimony at 1-4; Staff's TMI Testimony at 6; Tr. 1322).

CONCLUSIONS OF LAW

93. The Licensing Board has thoroughly reviewed and evaluated the evidence submitted by all parties with respect .

to the contention& raised. by the intervenors herein which are issues in this proceeding and with respect to the mat-ters raised by the Boa~d. 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-tion, as amended, the written testimony of all of the

wi:tnesses, the exhibits in evidence, as well as the., ,answers elicited from these witnes~es in response to questions of the Board and the parties and participants, the Board makes the following conclusions of law:

I (1) That there is.reasonable assurance that the activities authorized by the operating license amendment can

• • • • .be conducted. without endangering

.*

• the heal

. . ... th and safety*

of the public;.

(2) .That the activi.ti.es authorized by the. operating license amendment will be conducted in compliance

. with the.Commission's regulationsi (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 preparation of an environmental im-pact statement under the National Environmental Policy Act of 1969, as amended, 42 U.S.C. 4321, e t ~ - ,

and Part 51 of the Commission's regulatio:q.s, 10 CFR Part 51 and therefore that alternatives to the proposed action need not be con-sidered.

ORDER 9 4..

• Where£ ore, 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 herein, that the Director of Nuclear Reactor Regulation is authorized to make appropriate find-ings in accordance with the Commission's regulations and to

C i~sue 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.

§§2. 760, 2~ 762, ._2. 764*,

• 2. 785 and 2. 786, that this Initial

.Decision shall be ef~ec-:tive immediately and.shall constitute

• the final action of the Conµnission 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.

APPENDIX A - LIST OF EXHIBITS 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' 1-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 Letter f=o~ Librizzi to 3.5 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

(

Exhibit Mo. Descri?tion Transcript (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 prop.rietary 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

Exhibi+/--No. Description Transcript (Identified, Received 12 General Arranga~ent of 1338 1338 Auxiliary Building 13 Monticello Inspection 11 NRC at 339 Report, 4/10/79

..:*14*. *

• Licfensee rs Response to 165-2 Licensing Board Questions*

Regarding A "Gross Loss of

• . * .. * **wat:eri* from the Salem Spent Fuel Pool (unnumbered) Statement of Estimated Dose Submitted with Licensee's from Moving and Storing Response to Motion for Spent Fuel from Salem Unit Reconsideration of Cole-No. 1 in the Unit No. 2 man's Contention No.

Spent Fuel Pool and State- Thirteen ment of Fuel Elements Shipped from Salem Unit No.

l to Unit No. 2 (Attach-ments to Affidavit of Robert P. Douglas)

~

. ' ' .. 'CJNITED* STATES OF Af.!ERICA NUCLEAR REGULATORY CQ!,!MISS::i:8)7 Before the ~~crnic Safety a~~ Li2ansin~ Soard In the Matter.of J

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PUBLIC SERVICE ELECTRIC k~D GAS ) Docket No. 50-272 C0!-1.P.Al.'IY, et al. ) (Proposed Issuance of

) Amendment to Facility (Sai~ Nu'.-c:i~a:r .Gen~ra tipg **) operating.Lic~rise*

Station, Unit 1) ), No. DPR-70)

CERTIFICATE OF SERVICE I hereby certify that copies of "Licensee's Proposed Findings of Fact and Conc.lusions 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.~. Nuclear Regulatory U.S. NuclearI Regulatory

.

• I Commission *Cornmi s s 1. on Washington, D.C. 20555 Washington, D.C. 20555 Dr. W. Reed Johnson Dr. James C. Lamb, III Member, Atomic Safety and Member, Atomi~ Safety and Licensing Appeal Boa~d Licensing-Board Panel U.S. Nuclear Regulatory 313 Woodhaven Road Commission Chapel Hill, N.C. 27514 Washington, D.C. 20555 Chairman, Atomic Safety and Mr. Thomas S. Moore Licensing Appeal Board Panel Member, Atomic Safety and U.S. Nuclear Regulatory

~icensing Appeal Board Commission U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 Chairman, Atomic Safety and Licensing Board Panel 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

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\.. . 'i ...~.-

Richard Hluchan, Esq. Mr. Alfred C. _Coleman, Jr.

Deputy Attorney General ~rs. Eleanor G. Coleman Depa_rt.."Ilent of La*w and 35 "K" Drive Public Safety ?ennsville, ~ew Jersey 080.70 Environmental 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.sistan:t. Gene~al Sq-lici tor.

• Nor.thfield, N.J. 08225*.

Ji)ublic

• Service Electric:* .

• and Gas* Company Offic;:e o*f . the* Secretary

. 80 Pazk Pl-a*c-e '.: * .* * * . Docketing* and Service Section

. New~.rk , ...N ~ J ~ _.. *.o 7101 U.S. Nuclear* Regulatory

• Commission Raymond E. Makul, Esq. 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

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