ML23156A044

PRM-072-003 61FR24249 - Fawn Shillinglaw, Receipt of Petition for Rulemaking

ML23156A044
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Issue date:	05/14/1996
From:	Annette Vietti-Cook NRC/SECY
To:
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PRM-072-003, 61FR24249
Download: ML23156A044 (1)

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DOCUMENT DATE:

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ADAMS Template: SECY-067 05/14/1996 PRM-072-003 - 61 FR24249 - FAWN SHILLINGLAW; RECEIPT OF PETITION FOR RULEMAKING PRM-072-003 61 FR24249 RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

DOCKET NO. PRM-072-003 (61FR24249)

In the Matter of FAWN SHILLINGLAW; RECEIPT OF PETITION FOR RULEMAKI NG DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 03/14/96 05/08/96 05/22/96 07/10/96 07 /16/96 07/17/96

.07/17/96 07 /17 /96 07/22/96 07/22/96 12/09/95 05/08/96 03/05/96 07/09/96 07 /11/96 06/04/96 05/23/96 05/14/96 07/16/96 07/19/96 07/22/96 07/16/96 07/22/96 07/30/96 07/30/96 07/31/96 07 /17 /96 07/15/96 07/27/96 07/25/96 LETTERS DATED 12/9 AND 12/29/95 CONSTITUTING THE PETITION FEDERAL REGISTER NOTICE - RECEIPT OF PETITION FOR RULEMAKING LETTER TRAVERS TO SHILLINGLAW IN RESPONSE TO HER LETTERS OF 12/9 AND 12/29/95 RE SAR W/ATTACHMENTS MEMO TO DAVID L. MEYER FROM C. WILLIAM REAMER RE LTR FROM FAWN SHILLINGLAW DTD MAY 14, 1996 CCONCERNING PRM-72-3 COMMENT OF NEVADA NUCLEAR WASTE TASK FORCE, INCORPORATED (JUDY TREICHEL) (

1)

LTR TO MICHAEL HARRISON FM FAWN SHILLINGLAW LTR TO MICHAEL HARRISON FR FAWN SHILLINGLAW COMMENT OF FAWN SHILLINGLAW (

2)

COMMENT OF DIANA SALISBURY (

3)

COMMENT OF NUCLEAR ENERGY INSTITUTE (JOHN F. SCHMITT) (

4)

COMMENT OF ALICE H. HIRT (

5)

LTR FM DIANA SALISBURY TO SECRETARY RE 10 CFR PART 72, DOCKET NO. PRM-72-3, FAWN SHILLINGLAW (SUP TO COMMENT NO. 3)

COMMENT OF CHARLENE JOHNSTON (

6)

COMMENT OF CITIZENS' UTILITY BOARD (DENNIS DUMS) (

7)

COMMENT OF PRAIRIE ISLAND COALITION (GEORGE CROCKER) (

8)

DOCKET NO. PRM-072-003 (61FR24249)

DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 08/08/96 09/17/96 09/22/99 07/28/96 09/10/96 09/21/99 COMMENT OF MARY P. SINCLAIR (

9)

COMMENT OF NEVADA NUCLEAR WASTE TASK FORCE, INC.

(JUDY TREICHEL) (

10)

LTR FM SECY TO FAWN SHILLINGLAW ADVISING HER THAT ISSUES ONE, TWO AND THREE RAISED IN HER PETITION HAVE BEEN GRANTED; FOURTH ISSUE HAS BEEN DENIED

Ms. Fawn Shillinglaw 1952 Palisades Drive Appleton, Wisconsin 54915 UNITED STATES NUCLEAR REGULATOnY COMMISSION WASHINGTON, D.C. 20555-0001 September 21, 1999 Jv

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ADJ!._.,.

SUBJECT:

FINAL NRC ACTION ON PETITION FOR RULEMAKING PRM-72-3

Dear Ms. Shillinglaw:

In your letters addressed to Chairman Jackson dated December 9 and 29, 1995, you filed a petition for rulemaking requesting the Nuclear Regulatpry Commission (NRC) to amend its regulations regarding the control and use of safety analysis reports for spent fuel dry storage casks under 1 O CFR Part 72. On March 14, 1996, the petition was docketed by the NRC as PRM-72-3. In addition, in a letter dated April 15, 1996, you provided supplemental information on your petition. On May 14, 1996, the NRC published in the Federal Register a notice of receipt of your petition (61 FR 24249).

Specifically, you requested the NRC to amend its regulations that govern independent storage of spent nuclear fuel in dry storage casks to require that: (1) the safety analysis report (SAR) for a cask design fully conform with the associated NRC safety evaluation report (SER) and Certificate of Compliance (CoC) before NRC certification (i.e., approval) of the dry storage cask design; (2) the revision date and number of an SAR be specified whenever that report is referenc&u in documents; (3) the NRC must clarif~ the process for modification of an SAR after a cask has been certified; and (4) the NRC make the licensees' unloading procedures available to the public. Your supplemental letter recommended that to eliminate confusion, the term "CSAR" (i.e., cask safety analysis report) be used when referring to the SAR for any dry storage cask design which has been approved by the NRC and for which the NRC has issued a CoC.

The Commission has considered the issues you raised in your petition and has determined that issues (1), (2), and (3) above are being granted in part, and issue (4) is being denied. Each of the issues raised in your petition is addressed in a final rule amending 1 O CFR Parts 50 and 72, "Changes, Tests, and Experiments." The enclosed Federal Register notice on this final rule contains a discussion of the issues raised in your petition and the Commission's resolution of each of those issues (see subheading 0.2 "Petition for Rulemaking (PRM-72-3)"). This notice constitutes the Commission's final action on your petition.

Sincerely, Lv,=:.-~

Annette Vietti-Cook Docket: PRM-72-3

Enclosure:

Federal Register notice

NEVADA NUCLEAR WASTE TASK FORCE, PORATED Alamo Plaza 4550 W. Oakey Blvd.

'96 SEP 17 A_, :44 Suite 111 Las Vegas, NV 89102 702-248-11 27 FAX 702-248-11 28 800-227-9809 OFFICE OF ~r:: ~ETA RY DOCKETING & :r:- V

[ I Chairman Shirley A. Jackson U.S. Nuclear Regulatory Commission Mail Stop 0 -16 Gl5 Washington, D.C. 20555-0001

Dear Chairman Jackson:

September 10, 1996 DOCKET UMBER PETITION ULE PAM 7 2 -3

(&,I i:-R 2424~

The Nevada Nuclear Waste Task Force submitted comments to the Petition for Rulemaking, 10 CFR Part 72 - Docket No. PRM-72-3, Fawn Shillinglaw on July 11, 1996. We are very interested in the matters discussed in this rulemaking and are closely following the issue.

As an organization with the responsibility for disseminating information to Nevadans concerning nuclear waste, we have made every effort to gather as much material as possible. To do this, we interact with a variety of groups and individuals so that we can respond to inquiries and requests accurately and completely. With regard to the aforementioned rulemaking, we have received two publicly available pieces of correspondence from the Nuclear Energy Institute (NEI) to the Nuclear Regulatory Commission (NRC). This letter addresses concerns that we have about some of the statements made in that communication.

The first item that we find troubling is in the NEI comments to the petition. They request that response to the petition be held in abeyance pending the outcome of discussions between the industry and regulatory agency. They also state in this document that regulatory and technical challenges could cause dry cask storage delays and possibly impede plant operations. For Nevadans and concerned members of the public elsewhere, this is familiar and troubling language.

It implies that uninterrupted plant operation, referred to as "waste production" by some, takes precedence over matters of public health and safety. And the reference to industry and regulatory discussions suggests to many who already lack trust and confidence that matters of significant importance to their well-being are decided behind closed doors.

The other communication that raises public concerns is a letter dated July 1, 1996 to you from Joe F. Colvin ofNEI and an attached Analysis of Final Safety Analysis Report and 10 CFR 50.59 Implementation. I will point out disagreements that I have with the Analysis; but first, I am troubled by the statement in the cover letter... "This analysis is intended to serve as a basis for A

lSEP 2 O,~~II cknowleag~d by card...........................

• J.S. NUCLEAR REGULATORY COMMISSIOt-.

DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date __.4~VJ....,.Q'-------

Copies Reroived ____ l ______ _

.\dd'I Copies Reproduced _4 _____ _

Special Distribution Po R R t Q S Le.Sdc I C7a,1 I a'z1b er

further ( emphasis added) NRC and industry dialogue on this important topic to begin to address potential differences in understanding and expectations. We will proceed to interact with the NRC staff and Office of General Counsel on these matters." It is our belief that the NRC must recognize the seriousness of the current public apprehension concerning the interactions between the operators and regulators of nuclear facilities. At this time there is both strong citizen interest and objection to nuclear waste storage at reactor sites and possible centralized storage locations.

That concern and/or opposition is only increased when people fear that these important matters are being considered out of public view and with no citizen input. It also clearly shows the disparity between industry and public access to the regulator.

The NEI paper, Analysis ofFSAR and 10 CFR Part 50.59 Implementation does an excellent job of pointing out, in the FSAR history section, how licensing requirements have changed over the years. In what appears to this reader to be evidence that as commercial nuclear power plant operation became better understood, regulation became more thorough and comprehensive. It makes sense that the original requirements were, over time reevaluated and strengthened. That process is continuing as we learn more, as it should. In addition to better understanding the technology, the NRC is now dealing with issues related to plant aging and evidence that problems that could not be foreseen are now evident. For that reason, the insistence in this paper, that application of regulations should not become more stringent as well as the regulations themselves, is wrong. Even if, as NEI contends, FSARs should not have the force of law but should rather be reference documents, it would seem essential that they be required to be fully complete and accurate.

It appears, throughout the document, that NEI is seeking the lowest common denominator when it comes to licensing and operating requirements. They often refer to items of "safety significance." A wary public is uncomfortable with having the industry determine what that means and even more concerned when the situation is evaluated in terms of"economic justification." The case is repeatedly made that when the bare minimum of legal compliance is met--safety is achieved. That is not true and we have a growing number of examples such as the Millstone plant, VSC-24 casks, etc. as evidence.

The people of Nevada are extremely sensitive to these issues. They have been told for years that damage to public health and the environment from atmospheric nuclear weapons testing came as a result of a technology that was not fully understood. The continual message now is that we now know better. If this is the case, NRC is fully justified in requiring tighter adherence to existing regulations and full compliance with any new ones.

Faced with the possible siting of either a repository or an interim storage site, both of which are heavily opposed, Nevadans fear and distrust the government's assurances of safety.

They well know that currently the Department of Energy (DOE) is in the process of rewriting their repository siting guidelines just as NRC is writing new licensing regulations for such a 2

facility. They suspect, justified or not, that these steps are necessary to assure success of the program. No doubt both you and the DOE would disagree with that assessment but the public will have to be shown that they have misread the situation. The only way NRC can begin to have trust from a population that readily admits that they have become cynical, is to be tough, fair, and totally open to the public in the regulation of currently operating facilities. The evidence must be clear and undeniable that economic and political superiority does not allow greater access or relaxation of the rules for the commercial nuclear industry while the public is either kept out or only provided the right to speak, but not meaningful participation.

Agencies such as yours have different departments and staffs for the many issues you deal with. The public does not see them as separate. If problems arise at nuclear power plants, they assume that they can expect problems at any type of new facility. If rules appear to be unfair or too lax at one site, they expect that they will not be fully protected in their community.

When I attend meetings, hearings or conferences I am often asked, "How can we get people in Nevada to understand that the repository will not present a danger to them or that it interim storage is built, it will be safe?" Well, it's possible that you won't. But if it's true, as I have heard many times, that these facilities will never be built if so many people are so adamantly opposed, then serious steps should certainly be taken if any of these projects are to proceed.

Over the years DOE has undertaken a series of efforts aimed at building or restoring public trust and confidence. These essentially wound up being reports put on the shelf Nothing was ever done that included actions where the public could see get-tough, honest program policies where failure to comply with or meet standards stopped the process. Perhaps DOE was not in a position to provide such an example, but the NRC is. You can, through your oversight authority at Yucca Mountain, your regulatory role at commercial plants, and as the licensing agency for nuclear waste storage components, enforce and adopt stringent regulations that, if not met result in work stopping until compliance is proven. Such a demonstration is necessary, not just for the people of Nevada to rethink their stand on waste storage, but more immediately for the communities who now have NRC licensed facilities or where at-reactor storage is being proposed. I know of no region or population that has or will welcome these facilities where they live at the present time.

It is not NRC's job to keep projects moving, such as Yucca Mountain or to keep plants running as is alluded to in the NEI document you received. Your allegiance must be to the public.

Their health and safety must be protected to the level they determine is acceptable, without accommodating the economic health of the nuclear industry. The public should never have to worry that a facility will be granted a license just because the application was submitted or that operations will be allowed to continue if.all requirements have not been completely met.

3

You expressed those sentiments far better than I can in a speech that you gave recently in Connecticut regarding the Millstone plant when you said, "When you are talking about a health and safety agency... there should not be any question who the 'customer' is. The customer is the public. It is important that we do not make regulation any more burdensome or costly for the regulated industry than it needs to be, but the bottom line has to be that an NRC licensee operates a safe plant, and lives within regulatory requirements or it does not operate." That is all we ask.

Yours truly, el Executive Director 4

July 19. 1996 Mr. John C. Hoyle Secretary

~!:ofcwco US~RC NUClfAI (lflll INSTITUTE

'96 Jl 22 P J :5 7 OOCKETNUM&EA PETITION RlU fRII 7 2....

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IUDIOlOOICM. ~iff.

IMEAOENCV l'M:l>~.11 I~..an llll!~Tl(III U.S. Nuclear Regulatory Commieaion W~*hinston, D.C. 20556-0001 ATTENTION:

Dear Mr. Hoyle:

Docketing and Service B tanch Fawn Sbillinglaw, Recei,>t or Petition for Rulemaking (61 ~. Bu-242.C9-M:cy 14, 1996)

Reqve*t &t comment t Tbe Nuclear Energy Institute (NEl)1 often the followin1 oommente on behalf oCtbe commercial nuclear indw,try on the Fawn S ullinglaw. Receipt or Petition for Rulemaking. ** noticed in the F<<kral Iw1u rer dated May 14. 1996.

I)r,y etorage of IJ)ellt nuclear fuel is importa lt for the continued operation or nuclear power plante. Increased uae ofthi.a technol< gy will occur. Cballencee which are reewatoey and technical in nature could potentially cauee aipificant dry cask 1torap delaya and. in aome caeea, could iJbJ ~e p~nt operations.

Tbe petitioner re4uesi. that the NRC amen I ite regulations which 1:0vern independent etorap of spent fuel in dry cae~, etorage caeke to require that a safety analyai8 report (SAR) for a cask design fu1.l) conform* with the aaeociated NRC eafety evaluation report (SER) and oerti!ica*.e otcompliance prior to the NRC'e certification of the caek deaign. The petitio1,er also requeete that the revision date and identification number of the SAR be *P< cified whenever that report is re!ereMed iD docwnente.

1 Tbe N\dear Enero ln.tdtute (NE() ii the orsa.nb ation l"ffP()naible for enabliehins \IJ\ified Dudtar ind\llt.Q' poliC1 on matt.ere aft'itetine the nuc!ear enero induatry, inc:ludins the resulatory

~

of pneric operaUonal &Ad *hnicaJ iNue*. NEI'* membere include all utilitiee lice need t.o operate commercial n~ar power pl.ante in the Un.i

• eel StaLe1, nuclear plant deeisnen, major udu~enpneerift1 firm*. fuel fabrication lad.hue.. ma~riale liceneeee. and odw,r orgeniuit.ion.

and individl&ala invoJwd in the nuclear enero iP\du* t.ry.

9607250334 '160719

~3 PRH PDA L * ' I,;,-.. *.: I *~.. :,

'II' I

Mr. John C. Hoyle July 19, 1996 Pare2 Recently there bas been significant regulaw ~ activity regarding the licensing basis lor Cacilitiell and the statue of the SAR and SER in this regard. The practice of lllaintaining current documentation and the methods for doing eo also have been under consideration. These induetry and re(:ulatory discuaaiona could influence the merits ol this petition and i~ proper di8poeif ion. Enclosed (or your information is a copy or NEJ'e correspondence to Chairman Jackeon from Joe F. Colvin, President and Chie( Executive Officer, dated July 1, }!;96, forwarding an industry evaluation of the regulatory aignmc:anoe of the FSAR ar d related mattera. Pending the outcome ol these diecusaions, reeponse to th< -:.abject petition 1hould be held in abeyance.

We appredate the opportunity to comment o 1 the petition and look forward to diacu88ing this important matter with the NHC after the broader licensing baai8 i.Nuee currently under consideration have be m resolved. If you have any queetion.e or need lw-tber clarification, please call Alan Nelson at (202) 739-8110 or me at (202) 739 8108.

SiDoerely,

&.-: f-~

John F. Schmitt JFS/APN/ec Encloeure

IUCll&I £Nll5T IIStll,Jt July 1. 1996 Chairman Shirley A Jackson U. S. Nuclear RecuJatory Coznmission Mail Stop O*lG GlG Washington, DC 20555-0001

Dear Chairman Jackson:

The nuclear industry is moving forwar,' agg1.-essively to address a wide scope of NRC concerns related to license basis c<*nformance issues. The NEI Nuclear Strategic Issues Advisory Committee, C<*nsisting of each utility chief nuclear officer, ha~ been proactive in addressing these concerns and is considering initiatives for the industry to take to improve the NRC's confidence in each licensee's ability to operate in conformance with it., li~nsil> g basis.

A neceasary adjunct to these activities i*: to establish a clear UDdei.. tanding o!the regulatory framework that establiahes f1e licenmns basis &o eerve u a foundation for future activi~ea. In support of those activities, NEI prepared an evaluation of the regulatory siguwcance of the Final Ha!ety Analysis Report and related matters.

Enclo.~d is a copy of tbal analysis, entit ~.ed An4lysi.s of Fin.al So/ety Anal,.si.s Report and JO C.F.R. 50.59 lmpkrMIUGlion.

Because this analyaia necessarily addre{ sea a number of lepl ieaues, a copy ia beinc aent to the Office or General Couo* iel Wlder MJ)&rate cover *

'niia analysis is int.ended to aen-e as a bi Jlia for furtbtr NRC and iDdutry dialoeue on this important topic to begin to addre. sa p<MDtial differences in understanding and exp<<.-tations. We will proceed to int*,ract with the NRC staff and OGC on theee matters. Pleue contact me if you have I Q)' question* or if you think it would be beneficial to proceed in a difl'erent mann ?r.

Sincerely,

I

- --------------------~

• - *-* St Ct O 12011*ntl ANALl1,SIS OF FINAL SAFETY ANA).,YSJS REPORT AND JO C.P.R. 50.19 IM)>LEMENTATION June )998

J.

Introduction Two wue3 have arisen concerning licensee conformance to its Final Saf P.ty Analysis Report ("FSAR") and the proper UH of 10 C.F.R. 50.59 to make changes to the plant u deacribed in the FSAR. Durinr its revie*,., :.t these wues, aome members of the NRC *ta.fl' have stated that. in their view, a *1 commitments and atatements in the FSAR ahould be treated as *et.and*alone* re rw,rementa and therefore any difference between the FSAR and the plant would be i a violation of NRC requirements.

Unclerl)'ins thi., poaition aeema to be an aae unption that the FSAR ia a def.ailed compilation of a plant's licenain, basis.

Tbe.se developments would impact a long-st,mding resu)atciy echeme that is at the heart of the day*to-day operation ot nuclear power reactors. There may be apeci.6e wuu which require further industry and N RC di8cus.sion regarding licensee conformance with the FSAR and the applies tion of§ 50.69. However, no evidence has been cited that suggeata that the existi.o: regulatory syatem bu not ~

au~ in p,otectmr public health and at fety, without requiring conformance with narrow interpretations of each and eve. y statement in a licensee's FSAR, or that the diM:iplined process impoeed by I 50.59 has been renerically misapplied.

Tb.is paper analyzes the hiatorical evolution ofihe FSAR and the NRC'a treatment of the licensing basil from a legal pen,pectiVl! and the legal significance of the FSAR under current law, aod NRC regulati,,n.s. In addition, this p~r add.resaes recent NRC atafl' interpretation* of I S0.59 b,1 outlininc the dill'erences between the.ae iDterpretationa and lonptandi.ng indu.1 ;try practice. T>-~ intent of the paper ia to eatabliah a foundation upon which FSAR < onf'ormance ia.., and questions rerardinr the application of i 50.~9 can be e< oatructively n ecusaed by the NRC and the induatry.

~ a starting point, it is important to place ti e FSAR in appropriate context with respect to plant operational deci.aio111. 'lbe ei,t.ire reau)atory ayatem i.s hued on the premiee that such decisions should be (OCU8e{l on nuclear safety. Because of the complexity of nuclear planta and the comprel enaive regulatory aystem, there muat be aufficient flexibility to allow for rea.eoned <'edaiona in the day-io.day operation of the plant, conmtent with a lice!lsee'a obliga~on to comply,nth NRC regulations and ite license, includin1 it.a technical specifi, ations. NRC regulatiooa define the key reneric requiremente eatabliahed to pr<Mde reuonable usurance that public health and safety will be adequately protecte. I. and plant technical specifications define the key plant*srecific operational pan 11etera that must be met. The FSAR, on the other hand, contains analytical auppo1 t for these prescriptive operational constraint.a and general de9Criptions of the pl*1Dt and plant system.a, at.ructurea and component.3, which provided the buea upon, hicb the NRC fint. licenaed the facitty and eubse-quently u a reference docu,,ent for analyzing proposed cbanrea.

It has long been re<;0gu.iud by the age acy and the industrv that compliance with te<:h.ruca.l specification8 and rerulatior s ensure that the plant can be operated safely. The industry al.so believes thar licensees need to comply with FSAR commitmenu and with § 50.59 requ.ir< meots. This view, for example,~ reflected in the NRC-eodorsed NEI Commitment ~fan'lcement Guidelinea. 1 However, no evidence haa been publicly provided tb at suggests that the well-establi.thed enlorcemect mecbaoiams already i.o pJ ace to assure FSAR conformance and i 50.59 compliance are imuf!icient to address any problems that arise at individual licen.8M's or that otherwiH the emtin( regulatory acheme need be modified or the etatua oft.he FSAR redefined. ~ a rei-11lt, nonconformancea with repreeentations iJ:a the FSAR do oot neceaaarily tranala te into weiy iasues and do not neeea&arily mean that an underlying regulation or technical specification has been violated.

II.

FSAR: Backsrouod and Hist< rica.1 Perspective hi order to understand the lepl ligni.6 :ance ofthe FSAR as a bcensing dOCUlllent,

• an understandi.nJ of the hiatory of the development of the FSAR ~ neceuary, Orip.nally, 83 pan of the license applivition process, applicants prepared wba. was then called the 11bu&cb S\Unmary reJ){I rt: Upon hcensing. this hu.ards summary report ~e part of the licell8e. In 1~62, § 50.69 was promulgated. alone,nth a nviaed I 50.36. 27 Fed.~. 5491 (Ju, e 9, 1962). Tbe two rwM together allowed liceDHea to designate* portion oft.be buarda summary report u tecbni-=-1 specilicatioius. The technical specilicatz,na remained part of 1.he licenee, and Commission approval wa., required for any lice~ propoNd chanre. The hazards sunuaary report wa.s separated from th,* licenae, and could be changed by the licenaee in accordance with f 50.69. Tb at portion of the har.arda summary report that wu not part o£ the license/technict 1 specification.a becalne the FSAR as it is known today.2 Because of the historica~ role ofFSARa, they were not written with the precision oC reculationa. tecbnical ai <<:ificationa or licenae conditions. Rather, plant syetema, etruetures and compone, ta and design analyaes were summarily de<<ribed, and often refemKI to docume 1ta not part of the FSAR for more detailed diacuasiona.

1n 1968, I 50.36 was amended to define more clearly what should be included in the technical specifications. The Commiasic n intentionally hmited technical apecification.a to only tboee matters whi, h requi.N! "rigid conditions or limitations on

~

NEl M"morandum t.o NEI Admini3trtl,e Pointe ol'Cont.act re NEI Guidflline for Manatcmi NRC Commitment.a (January 30. 1006) 2 FSAlu WM*e pnmarily flummriN o(plant c!MJi<<n* and accident an11I~~ Early FSAR.11 **tn-very onmp.cl in C'Offlpariaon with ~nt f-'S,~

(e.x., two,,JumH v~nu* tod11y** twMty or mort volum~}

reactor opera Lion.* f2r1.land Qegeral ~lrtt, i' ~ompany. et al. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 271,74 (l97~J). These legally binding controla were denved £:rom analyse:- contaioed in t.he FSA R related to plant feature~ *that are of controlling importance to safety: l5l, citint Guide le Conwt! of Ttthnical Specifkotions for Nuclear Reactcrs, 33 Fed* Reg. at 18610 (November 1968).

Correspondingly, the FSAR did not have th,

• aa.ine legal import a.s when it had been part of the lic:enae, but inatead wu used to <lescribe the plant eyatem.t and analyaes that supported, amonr other tbinp, the tecl lD.ical apecificatioa.s. Plant chan,ee were controlled from a licen.sing pers:>ective by I 50.59, which included a requirement for lic:eueees to maintain recort' a of £.acility chanpa made under f 50.59 (i.t., chances to tbe plant or pl'OCt.:du -ea u d.aibed in the FSAR). 14.

Before the FSAR update rule (i.e., 10 C.F.R. 50. 7l(e)), records &MOciated with i 50.59 changes were maintained out.side,.hhe FSAR. Troian. 9 NRC at 271*74.

Althourh I 50.59 did require licenaeea to aul mit io the NRC deacripuona of changes, these descriptions were not requin i to be incorporated into the FSAR.

This meant that witil 1980, when I 50.7l(e) wu promulpted, FSARs were not requiNd to be maiJltained CWTeDt and conai.* tent with the u*built plant. Thus, prior to 1980, 1tatementa and commitments in the FSAR could not have had the tutus of beinr eta.ad-alone requirement., bee au.,e the plant would not have conformed to and wu not required to confon I to the FSAR.

When I 60.71(e) was promulgated in 1980, U e NRC stated that the purpose of havinr an updated FSAR was '"to provide an updAted reference document to be ueed in recuninr aa!ety anal7aea performed by t,b,

• licen"ff, the Commisai"D, and other intereated parties." 45 Fed. Reg. 30614 (May 9, 1980). The NRC further stated that submittal or updated FSAR pages did ns t constitute a licensing action, was not in~nd&d for the P\UJ>O&e ol re-reviewing pl8.D ts, and was only intended to provide information. In fact, it wu stated that the u1*date rule "is only a reporb.Dg requirement.* Isl at 30615. Section 50.71(e) did require licenaee& io update the FSAR to accurately reflect all analyae, aubm>tted to the NRC with the lieenae application, as part o!ihe licensing review pr >ee&S, as required by I 50.59 &Pd other NRC requirements, or as neoMSal'y to auppot ~ lioeDN amend.menta. Section

&0.71(e) did not reqllire any new aoalyau to l.e performed. lsL Further, I 60.71(e) did not cbanre the acope Clf t.he FSAR, although new information wu to be incorporated into the FSAR to relle-- t plant analyHes and plant changes that bad occurred since the initial FSAR was issued. For early plant.,, that ecope remained narrow; for more recent plants. tha( scope wu ah1*ady much broader.

From 1980 forward, FSARs as.sociated with n,1w planu rrew in sue and scope relative to older plant FSA.Rs, largely becau~c NRC requiremenui were evolving and powing, and addiuonal analy,"a and detaile<' plant descriptions were added to the FSAR in re$ponse. However, the legal l'f'~WJ'\ meat to develop a FSAR (i.e., to 8\lpport a L~nae appliCllltioo in accordan~ with§ 50.34(b)) was never chan~d nor 3

was the legal atatu8 of the FSAR ever fon,ally addrcJSed by the NRC. In keepinr with f 60. 7 l(e), the updated FSAR is now uaed by the NRC to perform variou8 li~IUing-re.lated reviews and by licenseet-to document plant information and important licensee commitments. Thia hf s ultimately led to greater NRC and licen.aee dependence on the FSAR as a ref,rence document, and still yet greater incluaion of plant information and detail, u.sociated with license commitments. In this.respect, the FSAR ia an import.ant do :u.ment but, as will be discussed later, that does not elevate the FSAR to a lepl.crtatus not otherwiM! established by law or recwation.

In summary. the hazards analysis report 'FSAR. once part of the license, was removed &om the license by the NRC in J 962..ua accordance with NRC regulations.

the plant could be chanced from the deao iption in the FSAR under § 50.59.

Becau.&e FSA.Rs were of varying scope an,l until 1980 were n~t required to be updated, they were not COEWBtent at any particular momeut with the as-built plant,.

or the aame from licenaee to licen8ee. Aft er 1980, FSARa were required to~

updated periodically in ao:orda.nce with J 50.71(e) to better support NRC and licensee analyse.s of lice0aing action.t, bu1 this change did not elevate the FSAR to the legal atature of a license condition. 'I Qe updating proceaa grew out of a proce&

that Ul',olved combining I 50.59 record., f nd other licensing documents into the oricinal FSAR to improve its value u a r d'ereace document. The level of detail in tbeae documents varied from licensee to Jicensee, and over time for each lioen&ee.

For theae reuona, updated FSAR.s were not oon4istent in their~ or level of detail.

Ill.

Legal Siplficance or the U pd, ted F'SAR Today The FSAR i8 an integral part of the li~1> se application and, in aecorda!lce with i 50.57, the NRC mu.st conclude prior to licensing that the facility baa been conatructed and will be operated in cont, rmance with the application, which includes the FSAR u modmed through the time of the NRC'a l'ffiew and iuuance o!the license. As diacussed in detail bel iw, specific prowdons in the updated !SAR are leplly eolorceable over the life of tb? plant in two respecta, that is, to the extent that a provision defines how com!,liece with NRC requiftmenta will Ix achieved and to the extent that a licena< e raila to chance or update the FSAR i *.

accordance with NRC rel\,tlations and/o* licensee procedurea. Tbe11e enforcemLi~

mechanisms ofl'er aubatantial recuJatol') control over FSAR conformance. However.

the updated FSAR is still not. in-and-or. itself,, legally binding document.

From a pru;edural perapective (i.e., the meclianum by which a change i.!I mal. *,, u a liceiuee fails to perform a required f 50 59 evaluatioo, the NRC bas treated ;Jat failure u a violation of§ l>0.59 and has wued Notices of Violation for such failureB.

The the-ory in cit:mc § 50.59 ia that a di! ferer~ between the plant and a rel*~

p~ion in the FSAR i~ equivalent to f plant "change" that. if not previously

~valuated in accordance with § 50.59, or.!inadequately evaluated, is a violation of f 50.59. In the NRC's view, this would bi true even if that aonconformance we~

unknown to the licensee and/or bad exi.st ~ 8ince initial plan 1. atart-up. The NRC hu historically taken this approach and apparently intend.! to continue this practice. Other potential "procedural" vi,,lations for differences between the plant and the FSAR could presumably, depend ng upon the nature and 900pe of the noncon!ormance, be based on 10 C.F.R. Part 50, Appendix B (e.g., Criterion Ill, Dai/in Control) and§ 50.71(e) (failure to update the FSAR). In certain circum.etaneea, a violation of I 60.9, Com, *ktouw and Accuracy of Information, could aleo be cited if the licensee £ails io 1,otify the NRC of information required to be complete and accurate in all material 1-e.spects ia not ao.

With ~t to the suhetantive upect of ticeo~ commitment8, there is an important hierarchical di8tinction that h, a been drawn by the NRC in the put:

commitmenta in the FSAR ue lerally biD ting if a failure to meet a commitment directly linked to an underlying reculato, y requirement - i.e., a regulation, license condition (including technical speci.6catio 1s), or order** results in a violation of that u.cderlying ~meut. OtberwiM! the failure to meet an FSAR commitment haa traditionally resulted only in a NotiC( of Deviations or, it applicable, a violation of one of the "procedural* requiremeot8 d, *acribed above. Further, the NRC has acknowledged that *a commitment ia not flll appropriate means to l'QO}ve an issue tbat bu a high aafety or reiu}atory aignif tcance... Such eipillicant matters are to be included either u conditions of the lia DH or u a part of the plant's technical specifications eo that they cannot be cb&.D~ without the prior approval of the etaff'." SECY-95*300, Nuclear EtU!rgy In,, iiuU's Guidan.ot Docu~n.t, -OuickliM for Managirig NRC CommitJMn.u.

• ThiJt NR'~ staff position on c:ommitmenu appropriately makes no distinction betwe-!n a commitment described in an FSAR or made in any other context.

There is legal precedent establishing the, *alidity ot this re,:watory approach. ~

fortln,nd Genet.al Electric Company. et al (Trojan Nuclear Plant), ALAB*531, 9 NRC 263. 272-74 (1979) (findinr that u :hnical specifications represent lepl bounda within which the liceosee i8 requi1 -ed io operate the facilit,): Long Island Liditiac Compuu (Shoreham Nuclear Pc,wer Station, Unit 1), ALAB-787, 20 NRC 1097. 112$-26 ( 1984) (holding that a licen.e Ctldition was not needed to enforce certain licensing commitments because th, NRC can enforce commitments in the FSAR by ord"r if necessary); and Ge;cutl ~ectric Companx (Wilmington, North Carolina Facility), DD-86*11, 24 NRC 325 (1986) (finding that eo!orceable requirements subject to a Notice ofViolat1on P.re "only requirement., specified in 3 NUREO-1600. *General Slfttf'm~nt of roli<'y a1 d Procedure for NRC Enfon:t-ment Artion,i:

(prt-Yk>Utly cod.Jied as 10 C F.R P*rt 2. Apper dUl C) provide. that NOlk'8 or De'\*iahon IIN'

• wnt~n noticN deecribin~ a uc-enl'M't fai.lure io utui(y a commitment where thf' commitment involved hae not~ mad~ a l~*lly bmdin11: rNJUi.rPTnent.* N"etion \1.D. R,loud A. d"' irw I NJ.tit.., A<'~

atatut.es, NRC regulations. license condi.tio1 i.S, or orders;.. other licensee commitmenu are enforcP.able by agency orc'er or by Notices of Deviation). Further support fox thi.9 i.Dterpretation is found in tJ,e ?\'RC Enforcement Policy, which defines a *requirement* as *a legally bin din: requirement such u a statute, regulation, license condi~on, technical spec:fication, or order;* and, as de8Cribed above, contruta that with commitments tb1 t have not been made "legally bindior requirement.,.* Section IV, n.5 and Section Vl.D. Also, the NRC Enforcement Manual, Rev. 1, Chapter 8, pp. 6-9 recopm,?S an enforcement hierarchy for lieenainr commitments deecribed in the FS/ Jt.

It appears that recent NRC ataft' statementf iun contrary to thu established line of lepl pretedent by sucgeating that all atateJ 1enta and commitments in the FSAR ahould be treated a.s *stand-alone* requirem mu such that a differenee between the plant and a related provi.,ion in the FSAR

• ould subject the been.see to the isauanoe of a Noti~ of Violation. Although there i8 no atatutory or regulatory proviaion that clearly supports this iDterpre'..ation, it has been iuggested that I 60.57(a)(2) could be read to elevat4! the lec-u status of tbP FSAR to that of eatabliahioc legally binding requirements.* The argumen1. ia that. becauff tbe FSAR is part of the licen.,e application, all p ~ons in it are legally binding in substance (i.e., become "at.and-alone* requir( meats) in the saJlle eente u a liceiue condition. Tbe contrary argument, however, u, that such an interprett.tion would be incon.ai.&tent with the NRC'a separation of the FSAR &om the tecl>nieal ep<<:ification8 and NRC practice reprding li* :enae ")Ommitments (e.g., NRC Enforcement Policy). 5 bi fact, if commitmen *;a in the FSAR are deemed to be li0en8e conditions, then a licensee could not change 1JUch a coaunitment other than in accordance 'Nitb I 50.90, and 5 50.59 would t e legally invalid.

Some licenM!'es have a general license condif on requiring operation of the facility in ar,eordawce with the liettnse application (whi< h includes the FSAR); some licenaeea have a specific license condition that explicify mention, the FSAR a.a the ducription or the facility being licensed; ra..no other licensees have no aucb provisions. Dependmc on the &xact wording or the hcenae condition, the FSAR or pa.rte thereof ~*Y be alleged to create a *spe< w* legll aigiuficance £or those licensees. However, it would be illogical to l'\*ad such a license condition as making lbe FSAR, RU & pan of' the liten3e.

Further evidence 8Upporti.nJ a conclU8ion th. t the FSAR, as a at.a.nd*alooe document, is not *Je~ally bindinr"' can be fou, 1d in the NRC technical specification Sf.ctK)fl 5-0 5'7(n)(2) at.attt that the NRC muH fmd that *tt)he fedlHy ""ill operate in eonformit~*

~1th the appu('at,on as amtnd<<-d. the pro't*uit00e of thr Act, and the rul~ and regullllion~ o( th, Comm illK>n. *

~ S~ch A1' irlterpret.at.on uf § 5-0.57(a)(2) woulrl be *r...w or difft-rent from a ~reva0\Jsly appucabl.-

mff' ~ihon* anci lhu~ wo\Jkt ronfttitu~" b~ltf1t ~ui.red to mK-t lhf' Nl'QUirementtt er( f r,o 109.

!wit/ill,~

r.

I improveme~t proiram. A fundamental pren ise of that program is that certain technical specification provisions should be, ~ocated to "licenaee-a>ntrolled documents... uch u the FSAR" because (1) th;tSe provwon! do not ri5e to the level of aiguificance that necessitates ma.king them pa.rt of the license like a technical epecificat.ion and, u a result. (2) they can be changed without prior NRC approval in accordance with§ 50.59. Final Policy StatelTWll 011 7echn.ical S~ficatwns Improvement& for Nuclear Power Reodors, 5t Fed. Reg. 39132, 39136 (July 22, 1993). The Commissfon a1ao stated in 1.bai P1Jliey Statement that *(c]ompliance with Technical Specificationa ~ required by tl.ie Commisail)n, and adherence to commitmenu, coniained in liceuee--controlle<' d'XUIDeDta ia expected." hL at 39138.

'Jb.i8 iDdicatea yet a,ain that the Commissioll recognized that the licensee*

controlled FSAR did not have the same lepl -,ianding as the liceD.9e. To now read all FSAR commitments as leplly bindinc in 1 he aame aeme u ih.e license would not only undermine the purposu and &hruat of the technical spec:ification improvement program, but would aleo appea, *.o contradict the distinction in the Polley Statement betweez, technical specifications and licensee-controlled commitments.

That havinc been said, there iJ, no *1ue!'-tion of what licenseca should do. They should ens1Ue that the FSAR ~ upda~ io at rve as an aQ;W"ate reference document, that changes being conmdered by, licensee io the plant or plant procedure~ ahou.ld be competently evalualed J*W'S\laDt to I 60.59, and that any differences between the plant confieuration a 1d an as&odated FSAR de9cription that have 8af ety i.mplicatiom are properly ide ~tmed.

IV.

Pan~* Current Lic,enains Dasi*

It aL,o mu.st be understood that the updated F SAR is 1121 a compilation of the current licen.sing basis. Durinr the Commiaai rm** deliberations l\lnporting the promulption of 10 C.F.R. Pa.rt M, &quiremt,u for Renewal of Operating Li<<MI!$

for Nuclear Power Platw, the NRC ~

that the current J;~sing basis

("CLB") w,u, contained in eeveral licensinc*re? 1ted documents, includinr the FSAR.

The NRC concluded that this approach, in cor-junction with exiatinr regulatory and administrative control proceuea, waa awlicie!,t io assure plant safety without requirinc a compilation af '1ie CLB into a a:i.D1 le, nand*alone document, or a new replatory proceu to control chuge,4 to tue C*~. Thia conclusion ia contrary to a poaition that the FSAR constitutes a detailed a>mpilation of a plant's licensing basis.

Tbe only re(\llation that addresses the concept of a CLB is Part 54. ln pertintint part, the CLB is denned there as *the set of N ~C requirements applicable to a specific plant and a licensee's 'ltr'ritten coaunib 1ent.8 for ensuring complianc~ with and operation within applicable NRC requi.re1 1e11ta and the plant*speci.6c design buui (including all modification, and additio1 s to such commitment.! over the life of I

the license) that are docketed and in effe ;t.... " 10 CY.R. § 54.3(a). 60 Fed. Reg.

22-tol, 22492 (May 8, 1995). The de.firut;oo further spt.cifiea that the CLB "includes the plant-specific design-ham., informati< o defined in 10 CFR 50.2 as documenud in~ most n~n.t final safety onolysis rejiort (FSAR) 1i.s required by 10 CFR 50.71 and the licensee's commitments remaini.r, r in effect tli.at were made in docketed licenainr correspondence such u liceose( r~onaes to NRC bulletins, generic letters, and enforcement actioWJ, as well hS licensee rommitments documented in NRC aafety evaluations or licensee event reporu.* 11.l. (emphasis added,).

The d.efi.n.ition of CLB in Pan 6.c ia,weep

• nc in that it includes, for the pUl'post,8 of UC'fd)..M renewal, licellaff commitment.a t.h lt arc not reculatory requirement.a and, tberelore, can be cha.og~ without NRC *1,proval (r.t., licenaee commitment., made i.u docketed liceoainc correapondence sucl-as lice11&ee responses to NRC bulleti.Ju, reoeric letters, and enforcement actions). Stt2 Cu,'ffnl Licensing Basis {or

~rating Planu, OPP-92-02, NRC 01lioe of Polir,y Planning, November 30, 1992, at 5. By its terms, however, the definitio>> of CUJ only include8 that portion or the FSAR containing plant*specific desirn inf,,rmatfon a., defined in 10 C.F.R. § 50.2 --

not the FSAR in it.a entirety. There is no f.iacu..sion in the supplementary information accompanyinc promulgation G f the final Part 54 regulations and/or the 1995 amendment& to the rule which indics iea that any portions of the FSAR beyond I 50.2 design basis information an conaid, -~. to be included in the 8COJ)e of the CLB. Stt 56 Fed. Reg. 64943 (Dec. 13, l~Jl): 60 Fed. ftes. 22461 (May 8, 1S95).

Thia, again, is incon.ai.8tent with the appa, mt aew NRC ataff'position that all proviaiona in the FSAR have the direct fol\~ of law.

Duri.nc the NRC'a delibentions on the CL), definition, the NRC clearly acknowledged that the FSAR did not cont,, in all of a licensee*, current licensing b~. ~

e.g., NRC SECY*94-066, Eval~ tion of Issues Discussed in SECY-92-814,

--Cummt Liceruin.6 &,i, for ~ratinl Pie,w* (March 15, 1994). The NRC further acknowled,ed that those commitment., th*: an out.side of the FSAR, but are pa.rt of the CLB, a.re not governed by any NRC reg i&.latory change and notification proceaa, but are controlled by licensee administr.&ti, *e Ptoee88e8 which have been generally effective in manacinc changes to the.e con, mitmenta and notifyinr the NRC when appropriate. liL The NRC found that exia~ inc NRC regulatory controls, coupled with lioen.eff *dminietrative processes, weJ e a,l.flicient to ensure that the CLB would be maintained to provide an accepta',le level of safety. ~l.(l at 11*13; att All& 59 Fed. Rec. 46574. 46577, 46582-C (September 9, 1994) (prop~

amendments to Part 64). Further, lor liceo ~ renewal, the NRC specifically rejected impoainc a requirement t.o "compile* the Cl,B into a !l'ingle document. 5.lt, e.g.,

NRC SECY-94-066; w &ls. 56 Fed..Reg.61943, 64952 (December 13, 1991).

Tbua, the NRC bas previously recognized tJ,at the F~AR does not contain all aspects or the CL~. that only portions of tlH

• FSAR are part of the CLB and that existinr NRC re5-llatory overairht and lice1 see ad.mini!trative processes, even if not consistent acr~ the industry, are adtq.aate to track, maintain, and change

commitment.e to U8ure.safety. This unde, cu'-b the notion that the FSAR bas been, or ehowd be in the future, a de1in.itive coo pilation of a plant's CLB.

Corre.,pondingly, an NRC focus on FSAR 1,onconformances should appropriately encourage Lcen8ees to update their FS~ to assure completeness and accuracy, but ahould A.21 require licenHe3 to compifr a detailed recitation of all licensing buia commitments.

V.

Section 60.G9 l88'Ue1

'nlere are two core isauea auociated with t~e recent positions articulated by memben or the NRC staff' regarding the a1 plication off 50.59, notwithstanding the long-etandin, practice related to these W\> ea.

The first is.sue is what co~titutea an unre, iewed safety question ("USQ") under the criterion related to increases in the probability of an accident or malfunction previously analyzed. Section 50.S9 states tbat an USQ results if the probability or conaequences ",.iay be increued.* In 1989, the indw,try completed preparation of NSAC 125, Guidelina for JO CFR 60.59 So/et~ Evaluatiora.s, to a&!i.st the industry in implementinc § 60.59 in an appropriate manner. Simply stated, NSAC-125 provides that am all probability incre~a d(

• not necessarily create a USQ, dependent upon on a safety evaluation oftJ,e effect oftheae increues. Where a chance in probability or consequence is eo,mall or the uncenainties in detel'fflinii:ig whether a change in probability bu OCCUJ1' -cl are such that it cannot be reasonably concluded that the probability bu actually changed (i.e., there is no clear tTeDd towards increaaing the probability), the ch, nee need not be considered an increue in prob~bility. However, a narrow interpre:.ation of this I 50.59 proruion would not allow Cor anx increase, even if'tbe increasf! is not significant as a matter of nuclear salety. For that reason, the NRC has not e1,do1'Md NSAC-125. However, N(eDt NRC gwcLuice6 provide8 that In considering the acceptability of a I censee*s 10 CFR 50.59 evaluation, the staff bu found comp< naating efrect.8 such u changes in administrative controb acceptable in ofl'aetting uucel'\aintiea and increues in the probability of oocum DC'" or consequences of an accident previously evaluated in the HAR or redllttioo!I in a margin of safety, provided the potential iPcreu ~s or reductions in margin are negligible. Normally. the detenninat'on of whether there is an increa&tt in the probability of occurre1 C" 4>r ronaequences of an accident previo\lsly evaluated in the SAR or a reduction in a margin of safety and whether such increa&e3 are negh;_;ible is based upon a qualitative 6

NRC lnsp<<-<:tton Manuel, Part 9900 JO CfR Gu danC'f., 10 CFR ~0.59, lnunm Gui~ 011 1)~ RNUir~m~nu Rd-m~ lo ChoJ\t *1 lo 1-tJC"illfit't, Proc-rdur,~ and r,,!$ (or F.'C{>i*rimr11 IJ.,~11U'd April !J, I 99G.

assessment ua:ing engineering eval,,ations consistent \lfith the original SAR analysis asaumptions. The co: npensatory actions must clearly outweigh any potential increase in probability of occurrenef! or consequences or reduction in margi 1.

The other major iaaue relates to safety ma~~- The NRC Staff has suggested that any facility or procedure change that coulc, result in a decrease in safety marcin. u analyzed in the FSAR, even if still within he pertinent regulatory requirement or eatablubed acceptance standard (t-6,, NU) lEG-0800), should be cowridered a USQ.

Howeve.r. lonc*standing indunry practice hu historically not conaid'!red small decre~a in the marcin between the resul.a of a prior analysia. a.a documented in the FSAR, for example, and the underlyinr rel'Ulatory criteria to be a USQ. Under either approach, the plant u modwed atilJ complied with the underlying lepl requirements thereby assurinr safety, heel use the reault remained more conservative than the regulatory requirem)Dt.

When reviewinc NSAC* 125, the NRC statt d that the aafety margin *ehould normally be considered the diil'erence be"' !ell the reeulatory limit ~

the limit speci.6ed by the reculations or Technical SJ <<:ificationa) and the value of the p11.rameter reviewed and approved by the a~aff u part of the licensing baaia for the plant... [which) ii typically the value of t1 e parameter proposed by the licenaee in the FSAR u modified by c.be staff's Saf'ety ~valuation Report(e). Tim value abowd be incorporated into the licen.eee*a updated FSAR and ii eometimee referred to u the 1ac:ioeptanc. limit."" The NRC abo state l that *(w)here a cbanp in mar,.iD ia to small or the uncertainties in determinini.._ hether a-cllanp in marpn bu occuned are such that it cannot be reuonably cond, 1ded that the margin hu actually chanced (i.e., there ia no clear trend toward reducing the marcin), the change need not be considered a rrouction in margin.* l~tter from Cbarlea E. Rossi (NRC) to Thomas E. Tipton (NUMJ..RC) dated May 10, 1989.

The NRC'a recent positiona in the NRC ln5} ection Manual Part 9900 ln.urim G~ on f 50.59 it\,plementation are in, onsistent with the NRC stafra previous position and the JUu!tinc industry practice The lmuim G~ states that "the marpn ahould be meuured apin.st that mi *.rpn deacribed in the facility U[Updated)FSAR, or the etafre SER iftbe lJFSAR doe. not describe the marpn, Thie dramatic chance in the buia agaiut.._ hich a proposed change ia to be measured bu profound implication,.

A question bu alao been raised rerardiD, tJ,e legal significance of NRC Safety Evaluation Reports ("SERa;. NRC SERa hf ve lepl significance in that they document the NRC stafra conclusions with 1espect to whether a facility proposed for licensing meets applicable statutory and re, Lllat.ory requiremeo~; they eatablish NRC positione against which the provwone ot S 50.109, &cltfitting, can be applied (m NUREO-1409, BocJctiuing Guideh~,. /.ppendix D, June 1990); and they can reiterate ai,nificant licenaee commitment~ (;tt, e.g., NRC Office Letter No. 34,

Utility Commitment,, July 31 1981). With respect to the application of§ 50.&9, however, the NRC ataied in its review of N~:AC-125 that an NRC SER "ia not auf&cient to conclude that implementation < f the modification does not involve a USQ because such SERs do not normally a~ dress the broader implications of a licensee'* proposal upon the facility u a wh >le.* att Letter &om Jose A Calvo (NRC) to Warren J. Hall (NUMARC) elated Jc'ebnaary 26, 1991; -~

Letter from Charles E. Rosai (NRC) to Warren J. Hall (NUMARC) dated December 26, 1991.

However, u noted above, the value ot the p1.rameter "'reviewed and approved by the stafl' u part ot the licensinr ba.m* i.a typical ty the va!ue of the parameter proposed by the licenMNt in the FSAR N wedifie:d by t'l;)c n,ro, SER. Tbw,, the exact lepl atatua or the SER (t.g., in implementatior.. of f 60.69) it not clear, mucb like the atatua of the FSAR subsequent to initial li(e.u:iDJ. However, in NSAC-125, the SER terves the role o£ estahliahing relevant acceptance standard., to aerve u benchmarks !or USQ detenniuationa.

VJ.

The Cumulative Effect Recent NRC etafr positions that would alter \.he existing FSAR / i 50.69 p?OCe$S could a1ao have major unintended efl'ecta. F< r example, a narrow interpretation of I 50.59 cowd require a licensee to seek a lice-1se amendment to make a plant cbanp or resolve an FSAR nonconformance uiat bu no impact on safety. Absent Cornmiasion relief while eucb a procedurally COCUHCI proceu is underway, a licensee may be torced to abut the plant down. Thi.8 c-:nald have,ap,;6cant couequencea for the public in terms of the reliable supply ore~ ectricity and the cost of replacement power, even though the matter has DR safety aipmeance. Thia augpsts that a rule or reaaon should apply in the application of§ 50.59 and the interpretation of the safety and leral sirnificance of statement, in the FSAR.

Ultimately, tbe re90lution ofwbat iJI in the public iDteresi will rest on the NRC policy deciaion o£ whether the NRC will con ti* 1ue its biaiorical focu.a on safety or ahi!t to a narrow interpretation or NRC reqw "ementa, 'llrithout rorard to Nfety aisnificance. Thie decision will have profoun,\ implications to the allocation of both NRC and licensee reaourcea.

VII. Conclusion Based on the foregoing, an me stafl'positioD that all license commitments and/or etatemenu in the FSAR ahould be treated as directly enforceable "stand-alone" requiremenl8 and, therefore, any departure f> om an FSAR provision would constitute a violation of NRC requirements, d >es not appear to be legally support.able, nor consistent with the agency's long*1tanding practice, or justified by any change in circumstances. The historical <levelopment of the FSAR and the NRC'e historic treatment or the FSAR do not,.upport the proposition th Rt the FSAR 11

i, a lerally binding pa.rt of the license upo 1 which edoreemeot action can be directly based. Moreover, there*are well*e 1t.ablisbed enforcement mechanism&

already in place to ensure that licensees c,,nform with applicable comZIUtmenta in the FSAR, (i.e., those "procedural" regul.at ons which control licensee changes to the plant and FSAR), and underlying legal re,juirementa (i.e., regulations, orders, and hceiae conditions, inc:1udinr technical ape i6cations) to the extent tbd the failure to meet a specific commitment in the FSAlt results in a failure to comply with that widerlyi.ng requirement. There ia z,o neet. to create a new lepl aipmcance for the FSAR.

Additionally, suggeatio~ that the FSAR s':iould have been, or ehould be in the future, the definitive detailed compilation of... plant's licensing b~. appear to be inconaiatent with pr~or NRC staff' auea.9m ent.s of the lack of a demonstrable benefit of compiling a CLB. The staff has alread}' acknowledpd that the CLB, spread over several documents including the FSAR. a,,d controlled by ewtiDg NRC regulatory and licensoo Bdroinistrative proceMea, is E u.fficient to eruiure plant safety. No change in circumstances hu been identifi ?d that would justily a reversal of that conclusion.

Recent FSAR nonconfonnances found at~ >me plants do not~ safety wuea of auch significance that the licensing procei s lor the entire indu.etry aho'1.ld be subject to the major upset that will re8ult if the N RC si.aff continues on its appuent path of modifying in a significant manner, a lone* standing regulatory echeme, particularly in the absence of evidence that an econou, ically justified subatantial increase in protection of the public health and aafety would result~ I 60.109). Undermini1>g the eat.abliahed reru}atory procesa could b ave several conaequenoes (e.g.,

unneeeasary plant shutdown.a) that could have a aig'ni&cant economic impact on the public and the industry with no safety be;,efit. There is absolutely no question that isaues of aafety significance have been, az d can continue to be, forthrightly addressed under current regulatory procE "9e8 and practice&. 'niere may be epeci.fic uauee nprding conformance with the ~.AR and the proper application or I 60.59 that need to be addreseed. However, tbe1 e ia no need for the reculatory framework to be modified throup new interpretatio, -a of long-standinc requir-Nnents and practices.

12

V ~!{ET NUMBER

~-ETITION RULE PRM '7:2. -3 G6 -t F-, 12.,:)_t-j Z.~9) 5711 Summerset IOOCKET f-"0 Midland, MI 48640 USfJ RC July 28, 1996 Secretary U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Sir:

I wish to submit the following comments on dated May 14, 1996, vol. 61, No. 94, p. 24249-50 rulemaking submitted by Fawn Shilling law.

'96 AUG -8 P 2 :59 OFFI CE OF SE r~ETr.xR Y OOCKE.1 UC:i &.

-K ii Cf.

BRiJ, '"~

the Federal Register notice regarding the petition for (j)

I agree with the basic premise of the rulemaking request that the present methods of making changes in casks that are to store high level nuclear wastes on the shores of the major fresh water supplies of the nation and the manner of reporting them are inadequate. They can lead to a situation where the SAR does not conform to NRC's SER and COC.

This can jeopardize public health and safety.

The many changes made in casks after they have been certified as "generic" by the Nuclear Regulatory Commission has made this rulemaking necessary. This chain of events demonstrates that the term "generic" as the NRC is applying it in the licensing of casks for dry storage of high level nuclear is meaningless in any real sense of the word.

The NRC should vacate the "generic" ruling procedure and require all site-specific studies and needed changes at various nuclear plant sites to be subject to a public hearing prior to certifying a cask in order to avoid the kind of confusion that now exists in the entire dry cask storage process.

A good example of this confusion from the failure to properly document changes in cask design-on the part of workers, management and the NRC-occurred during the flash/fire at the Point Beach plant on May 28.

This hydrogen gas fire was of sufficient explosive force to lift and displace the shield lid which was actually 6390 lbs., two tons more than was first reported.

This happened because WEPCO used 72.48 to change their shield lid design yet did not make changes to reflect this new weight in their documents.

As a consequence all the calculations for this heavy load in their procedures were erroneous, based on numbers for the weight of the shield lid prior to the welding of the original 2 parts to the shield lid into one lid weighing far more.

This lid at Point Beach was also altered to include a larger vent than the one that was originally certified as "generic" at Palisades.

This change was made to improve the unloading process.

Consumers Power Co. has also changed this lid design from two parts to one since the initial certification of the VSC-24 cask at the Palisades site.

Yet kknowlect ed b

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. _, l:lCLEAR RE ~1L,:..i-."iOHY (;OMNilSSIO OOCK~TiNG & 3ERViCE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statlsb Postmark Date "1 /J Q/ 0 [

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6

this one part design remains in the YSC-24 document as a two part design. (p. 1-

9)

The SAR is the basic document that should contain all the details of cask design and its use.

It is the document that the public commented on in the ru1emaking procedure for the VSC-24 cask.

Therefore, any changes in design should be made in this document through an amendment allowing for public comment.

The SAR must be viewed as an integral part of the certifying documents, along with NRC's SER and COC, for any cask design.

In order to dispel the confusion that now exists in cask documentation, it is essential that the NRC should not certify a cask until the SAR reflects all the SER and COC requirements.

As changes actually were made in the VSC-24 cask SAR after it was certified, revisions OA and OAA were needed to satisfy NRC "generic" conditions after the casks were already loaded.

The NRC owes the public an explanation as to why, and under what regulatory procedure, is the vendor, who has never been licensed in any way by the NRC, allowed to make changes in the SAR after the NRC has certified a cask.

I also agree with the petitioner that the unloading procedures should be described in detail by the vendor before a cask design is certified.

This document should be available in the public documents room before any cask is loaded.

Any changes that are made should be done by amendment to the SAR and allow for public comment.

Any changes or evaluations by utilities under 72.48 or 50.59 should also be put in the public documents room.

At the present time. the NRC is still reviewing and pondering approval of unloading procedures for the untested VSC-24 casks at Palisades a year after 13 casks were already loaded with high level nuclear waste 150 yards from the shore of Lake Michigan.

This demonstrates NRC's failure to place public health and safety above the cost cutting priorities of the utility.

There is an urgent need for the NRC to call for a moratorium on the dry cask storage process until responsible procedures for documentation for the design, use and unloading of casks are implemented which not only the fabricators, vendors and licensees, but the public can rely on as being accurate and dependable.

P.O. Box 174

• Lake Elmo. MN 55042

• Phone: 612-770-3861

• FAX 770-3976 Petition for Rulemaking Nuclear Regulatory commission 10 CFR Part 72 (Docket No. PRM-72-3)

Fawn Shillinglaw

• 96 JUL 31 P 6 :os DOCKET N1J~v:BER July 25, 1996 PETITION C..,...iLE PRM ?2-3 Petitioner Fawn Shillinglaw requests that the NRC amend its {'=>ll=R 242.fi regulations governing dry storage cask certification so that the V

safety analysis report (SAR) for the cask design fully conform with the NRC's safety evaluation report (SER) and certificate of compliance (COC) prior to NRC certification of cask designs.

The petitioner also requests that 10 CFR Part 72 be amended to require that the revision date and number of an SAR be specified Whenever that report is referenced in documents, and that the modification process for a SAR be clarified.

Finally, petitioner recommends that the NRC make cask unloading procedures publicly available.

The Prairie Island Coalition submits these comments in conjunction with its concerns about nuclear operations, including nuclear waste management, at Northern States Power co.'s (NSP) Prairie Island plant and at other commercial reactor and nuclear waste storage sites.

The Prairie Island Coalition supports the recommendations of the petitioner, and further requests that NRC require a full rulemaking amendment procedure with a public comment period for any change or revision made to the SAR.

The need for petitioner's recommended changes, revisions and disclosures is evident and obvious in light of the welding problems at Palisades and the explosion at Point Beach, both involving VSC-24 casks.

Were the recommended SAR revisions in place prior to the loading of these casks with irradiated fuel, the risks and threats posed by these problems and incidents are not likely to have occured, as the VSC-24 cask design is not likely to have been licensed.

The Prairie Island Coalition is particularly concerned about the lack of tested and approved cask unloading procedures.

Already, in a very short period of time, unforeseen circumstances have arisen afte~

a licensed cask has been put in use that require the cask to be unloaded.

As cask storage operations increase, and bearing in mind the numerous violations of cask fabrication protocol experienced during fabrication of the first TN-40 storage casks in use at Prairie Island, it is virtually certain that additional similar unforeseen circumstances will occur.

In this setting, the lack of tested and approved cask unloading procedures creates an unacceptable risk of massive, uncontrolled radiation releases.

Our concern about problems with unloading casks is amplified by the NRC's staff "BRIEFING ON STATUS OF DRY CASK STORAGE ISSUES" on May 30, 1996.

At No. 47 of ~he transcript from this public meeting,

• 1.s. NUCLEAR REGUI.ATOfW COMMISSIOh DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISS,c)N Docootent Statistics Postmark Date 7 b. (tJ fci~

Co~esReceived_1_1_ ',.-___ _

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J...csa1/2 qallt3b ec

Andrew Kugler, Lead Project Manager, Dry cask storage, NRR, states, "For the unloading procedures, what we are finding is that they are more complex than the loading procedures.

Unfortunately, some of the older SARs fail to recognize this and tend to indicate that unloading is simply the reverse of loading, which is not true."

Mr. Kugler goes on to identify cask unloading concerns regarding the condition of the fuel after being in the cask for a period of time, cask pressurization problems due to steam generation, thermal shock during reflooding, radiological protection, and the total lack of any cask unloading experience.

One "older SAR" failing to recognize that cask unloading is not simply the reverse of loading is NSP's license for TN-40 casks at Prairie Island.

(See Attachment)

As part of a full rulemaking amendment procedure to address the issues raised by Docket No. PRM 3, and to initiate a safe resolution of the contradiction between NSP's ISFSI SER and the May 30 comments of Mr. Kugler, the Prairie Island Coalition requests that all cask loading at commercial reactor sites be suspended pending fully tested and approved unloading procedures.

Without such procedures, both worker and public health and safety are not adequately protected against the clear and present dangers of radioactive contamination created by the potential need to unload a failing cask, and by the absolute need to unload a cask at the end of its useful life.

The NRC is charged with the responsibility of preventing radioactive contamination of workers, the public and the environment.

To do so, the NRC must have the confidence of citizens.

This requires broad public input and involvement in rulemaking amendment procedures that are impeccably thorough, complete, and fair, and that result in management requirements that actually will prevent radioactive contamination

• As existing nuclear waste management requirements create opportunities for confusion, error and accidents, and by the NRC's own admission do not adequately account for the complexity of the situation, the Prairie Island Coalition fully supports the Petition for Rulemaking of Fawn Shillinglaw.

The Prairie Island Coalition further requests that the necessary changes, including the suspension of all commercial nuclear waste cask loading activities until such time as unloading procedures are tested and approved, be made by way of a full rulemaking procedure with a public comment period.

Sincerely,

.J?'.e,,c-,-~

George Crocker, Steering Committee Prairie Island Coalition

FROM: Panasonic FAX SYSTEM PHOl'-lE NO.

Mar. 06 1995 03:03AM Pl SAFETY EVALUATION REPORT FOR THE PRAIRIE ISLA:'\1>

1DEPE!DENTSPL~T Ft:EL STORAGE INSTALLATIOS tr.S. Nucl~ Regulatory Commission Office of Nuc:lear.Material Safety and Safeguards

• . Jul\'. 1993 **

~

9311010119 931019 PPR

• AQQ.Ck 05000282 Enclosure

FROM Panasonic FAX SYSTEM PHONE NO.

Mar. 06 1'395 DJ: 03RM P2 I.?.

DECOM.'\flSSIOSl!'-iG Cask decommissioning is deemed acceptable,fit can be shown that the applic.able regulations h.ave been folto1A*ed. as appropriate. In addition. -.*here limits can be applied. these limits have not ~en e,ceeded.

U *.?

.\~ct'ptanc:~ Criteria

• . 10 CFR 7Z.30 provzdes requirements tor a site~specific decommissioning plan. including financin~. Among the items to be addressed under. this pan are the decontamination of the site ancJ facil,ues. the di~posal of rcs1dual radioacth*c materials atter all spent fuel has been remO\*ed.

the decomm1ssion1ng funding plan. and the cost estimate for decommissioning.

10 CFR 'n.130 provides requirements for decommissioning and states. in part. that the ISFSI shall be dcs.gned for decommissioning. Among the* items to be addressed under this pan are the provisions to facllitatc decontamination of equipment. the provisions to minimize the quantity of radic.1ac:fr11*e wastes and contaminated equipment. and the provi5ions to facilitate the removal of radioacn"'e....,.astes and materials at the time of permanent decommissioning.

-N CFR 173.411. 173.4:3. and 173.435 provide information on the radionuclide acti\*itics that may be transported as limited quantity materials.

10 CFR 30.14 and JO. 70 address radionuclide concentrations th.at are exempt from licensing requirements.

10 CFR JO. l 8 and )0. 71 address rad1ooudide quantities that are exempt from licensing requirements.

10 CFR 61.55 and 61.56 address the radac'.>nuclide concentrations for Class A wastes and the characteristics of such waste.

The review is divided into four ma.in paru: (1) unloading of the cask. (2) decommissioning of the cask components. (3) decomm1ssioning ~f the storage pad. and (4) decommissioning funding.

1.2.J. I Unloading or the Cask I

The possibility for cask unloading is mentioned in Section 4.6 ('"Decommissioning Plan*) of the TSSAR. NSP has provided supplemenr.ahnfonnation. on the wet unloading (in the spent fuel prol) proci:durc. in a response co qucstio!'ls:. datfd July IO. 1992 (Reterence 26).

12-1

FROM Panasonic FA>< '3'r'STEM Mar. 06 1'395 03: 04AM P3

. ~

Assuming a normal spent fuel pool transfer. the unloading sequence is essentialJ)* a reverse of

---

,,, the loadinc sequence. The cask is mumed from the ISFSJ to the auxiJiary building rail bay via the transporter. The weather cover is unbolted and removed. The O\'erpressure system is then relN>'led and the cavicy gas sampled thn ugh the van pon. Once the cask is moved into the f ud pool area. the cavity is depressurized and the ask Jowered into the spent fuel pool. With the cask lid at the pool surf acc. fill and drain lines are connected to the lid drain and vent ports.

Borated water is slowly added to fill the cask and gradually cool the fuel in the cask. When the cask is full. the fill and drain lines~ removed. The cask is then lowered to the pool bottom.

where lhe lid is removed. making the fuel accHsible for tiansfer.

Decontamination of the cask is addressed in Section 4.4.l of the TSSAR. The text of this section states that '"Standard decontamination methods will be wed to remove surfate conramination from the asks.* It is further stated that "Manual methods will be employed, using water detergents and wiping cloths.*

• Decontamination of the cask is addressed in Section

4. 6 o( the TSSAR. The text of this xction mentions surface decontamination using chemical etching and/or ~lectropolishing. These methods are accept.able to the staff. provided that detailed procedures arc developed before decontamination of a cask. '

12.J.l Decommissionina of the Cask Components Occommissioning-of the neutron activated fuel basket. cask body and lid. neutron shield. and neutron shield shell and protective cover is addressed in Section 4.6 c*Dccommissioning Plan")

of the TSSAR. Supplemental information is also provided by NSP in responses to questions dated June,. 1991. and December 23. 1991.

N~utron fluxes obtained from the XSORNPM shielding calculations arc used by TN in conjunction with ORIGEN2. to calculate the activation products at 30 days subsequent to unloading. Only those nuclides with activity greater than.37 MBq ( 10*5 Ci) and those nuclides listed in 10 CFR 61.5S are reported in the TSSAR.

Table 4.6-2 c*Results of ORIGEN2 Activation Calculations"') summarizes the activities for s*cr.,.Mn. "Fe.,..Fe. "Co *.OCo. *Ni.

and 4tNi in the fuel bask~; '"C. Cr.,.Mn. "Fe. "Fe. -Co. -Co. "Ni, and ""Ni in the cask body and lid: 'H, *~. and.,Zn in the neutron shield; and *4(: and "Fe in the neutron shield shell and protective cover. Some of these nuclides emir no gamma rays ('H. C. and "Ni). Others ('1Fc and "Ni) emit gamma rays with energies of less than 8 keV.

Materials quantities used by TN in the activatial' calculations are representative of those in the cask itself. Weights for the fuel basket. cask body and lid. n~tron shield. and neutron shield shell and protective cover are summarized in Table 4.6-1 c*Dara for TN-40 Activation Analysis") o( the TSSAR and are 5676 kg (t:2.513 lb), 69,465 kg (153,144 lb), 5766 kg (12,712 lb). and 4 D 1 kg (9107 lb). respectively; volumes for the fuel basket, cask body and lid. neutron shield. and neutron shield shell *and protective cover are summarized in Table 4.6--3

(*Comparison of TN*40 Activity with Class A Waste Limits*) of the TSSAR and are 1.43 m'

(~0.5 ft'). 9.27 m, (328 ft>). 3.37 m' 019 rt>). and 0~53 m, (18.7 ft'). respectively.

I

FROM:.Panasonic FAX SYSTEM PHONE: t-HJ.

Mar. 06 1995 03:0SAM P4 In evaluating the activation products. the :s.~ff has assumed the activities. weights, and volumes p~nted an TSSAR Tables 4.6-Z. 4.6-l. illld 4.6-3. respectively. Thiny days after the removal or the fuel assemblies. the following 1m,cn1oncs remain in the principal cask components.

INVENTORY (MBq)

Fuel

• Cask Body Outer Shell and hc:nope Bask~

• arid Lid Neutron Shield Protective Cover

'H 7.8,hl0*

14c

. 7.47"1~

L89xto*'.

l.:?9~ 10 *

~'Ct 141 13.S "Mn 17.1 309 HfC 187 J360

.SSJ

~.. Fe J.46 6:::.5 uCo

~ 1.8

'\.,.,

.,.,Co

.J08

.307

'"Ni

.117

.119

~'Ni

• 13.S I :

13.8 4'Zn

.414 12.J.J Dttummi~4-ioninc or th* StOl"ll&t! Pad Decommjs.~ioning of the neutron activated sto~ge pad and underlying media is addressed in NSP r~pon ses to questions dated J unc S. l 99 l

• and D<<erriber 23. 1991. Calculations of the neutron act1vation of the storage pad and underlying mcdia'are not performed. Neutron nuJEes at the bottom of the cask arc estimated to be approximately an order o( magnitude less than that used in the cask body activation calculations. With the cask activation product activities as a basis.

NSP infers that activation of the storage pad.and underlying media is less than that of the cask and. therefore. below the altowahle limits for Class A waste. Specific procedures for disposal of the stora,e pad remnant! and underlying media are not pro\"ided.

12.3.4 0-Commmtonina Fundlnc :,

Decommissioning funding is addressed, in Section tci I (*Finandal Qualification*) of the NSP

.. Application for a License to Construct and Operate a Dry Cask tSFSI.

• Decommissioning costs are currently estimated to be $3,100.000 and will he added to those for Prairie Island Unit

l. Financial assurance for JSFSt decommissioning costs is provided by the external sinking fund.

with monthly deposit~. established for plant decommissioning. Beginning in 1993, the site*

~pccitic cost estimate will be revised on a, 3*y~ basis by performing a new analysis. The current costs wm be adjusted by inflationi to determine the finaJ cost. before the recovery pattern is determined for the external sinking fund.

12-l I

,*:.. ~

FROM Panasonic FAX SYSTEM PHONE t'-lO.

Mar. 06 1995 03:OSRM PS 12 *.a F"mdinp and Conclusions Cask unloading information is consistent "":1th the regulatory requirements.

Activation product concentrations usociated with decommissioning of the cask components att

\uch that the cask components contain li~~-exempt concentrations of 'H. iic. ~*er. s.Mn.

~\Fe.,,.Fe. '"Co. --*co. and "~Zn. Funhcrmo~. the activities or concentrations are such that the cask componenrs may be classed a~ limited quanmy materials for off she transportation and ma~

be dt'ipc'Xed of as Class A waste.

• Oecommi~sioning funding inf("rmation is consistent wirh the ~~gulatory requirements.

The cask design ts consistent with the requirements of 10 CFR 72.130 that an ISFSI be designed for decommissioning. Funhermore. the action$ involved in cask disposal and decommissioning fonding arc cons"tent with the requirements oi 10 CFR 7~.JO as t'ea~ibl~ eh:ments of a site-.

~p<<ifk dccomm,uioning plan.

. I' 1;

16 N. Carroll St., Suite 300 Madison. WI 53703 (608) 251-3322 CITIZENS' UTILIT't°~ARD July 27, 1996 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 Re:

Petition for Rulemaking 10 CFR Part 72 Docket No. PRM-72-3 Fawn Shillinglaw

'96 JUL 30 P 5 :06 G) ooci<ET NUMBER FETITION RULE PRM 72-3

( bl F ~ 2 4 2. 4CJ)

Enclosed are the comments of Citizens' Utility Board to the above referenced petition for rulemaking.

Sincerely, Dennis Dums Research Director cc:

Fawn Shillinglaw AUG 1 3 199q Acknowledged by card -...... _*-*"*"*--.. w-..

JJ.S. NUCLEAR REGULATORY COMMISSIO~

DOCKETING & SERVICE SECTk)N OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date 7/ 27 / '16' Copies Received._

..._l -

.\deft Copies Reproduced =-1~ ---

Speclal Distribution PDR RI OS Lesar ( (iallajbec

• Petition for Rulemaking 10 CFR Part 72 Docket No. PRM-72-3 Fawn Shillinglaw The Nuclear Regulatory Commission (NRC) has interpreted statements made by Mrs. Shillinglaw in letters dated December 9, 1995 and December 29, 1995 to Chairman Jackson as a petition for rulemaking.

In Mrs.

Shillinglaw's December 9th letter she requested the following actions, asked the following questions and made the following relevant comments:

"I request that when a generic cask SAR is referenced in any documents, the Revision number is supplied and the date as well." (Fawn Shillinglaw letter to Chairman Jackson, December 9, 1995, p 1)

"Ever since the proposed rule for the 1st generic cask certification I have requested a final SAR be accepted before a cask is certified and I still think that should be required.

The SAR should completely fulfill all requirements of the NRC SER and C of C before it is issued to anybody to use. (Fawn Shillinglaw letter to Chairman Jackson, December 9, 1995, p.

2-3)

"Now that we have SAR OAA, how is that document allowed to be changed in the future? It is my understanding that to change this document, rulemaking and an amendment is required, as the cask was certified (SER, SAR, CoC) by rulemaking.... What is the process for a vendor to change the SAR generic document issued to a vendor cask design?

It is an amendment done by rulemaking." ( Fawn Shillinglaw letter to Chairman Jackson, December 9, 1995, p 3)

In Mrs. Shillinglaw's December 29th letter she asked the following questions and made the following relevant comments:

"What process did NRC use for reviewing and allowing Sierra to make changes?... Why was this done without the proper procedure of amendment and rulemaking?"

( Fawn Shillinglaw letter to Chairman Jackson, December 29, 1995, p 1)

"NRC secrecy in cask unloading procedures will undermine public trust in any dry cask storage program.... "

( Fawn Shillinglaw letter to Chairman Jackson, December 29, 1995, p

2) 1

The NRC's summary regarding Docket No. PRM-72-3 appearing in the Federal Register at Vol. 61, No. 94, p 24249 requests comments on the three following issues:

1)

Should it be required that the SAR for a cask design fully conform with the associated NRC SER and COC before NRC certifies the cask design?;

2)

Should the revision date and number of a SAR be specified whenever that report is referenced in documents?;

3)

Should the process for modification of a SAR after a cask has been certified be clarified?

Issues 1, 2 The NRC should require that the SAR for a cask design fully conform with the associated SER and COC before it approves of the cask design, and as well, the NRC should require that the revision date and number of a SAR be specified whenever that report is referenced in vendor, licensee, and NRC documents.

These actions would reduce the possibility of there being contradictions and differences between the VSC-24 cask SAR and the NRC SER and COC, and would reduce confusion regarding what documents form the licensing basis for the cask.

Issue 3 A review of Mrs. Shillinglaw's letters to Chairman Jackson makes it clear that Mrs. Shillinglaw was concerned that Sierra Nuclear made changes to the design of the VSC-24 cask outside of the NRC's rulemaking process, which is the only way an SAR can be amended. Mrs. Shillinglaw explicitly asked for a response from the NRC as to why Sierra Nuclear was allowed to change the design of the VSC-24 without the NRC approving of the changes in a rulemaking procedure.

Mrs. Shillinglaw did not request that a different procedure be put into place to allow a cask vendor to change the design of a cask, or that the NRC should consider putting a different process in place.

It would be inappropriate for the NRC to utilize an inquiry of such a dedicated citizen as Mrs. Shillinglaw to revise regulations to allow a cask vendor to change the design of a generic cask in any way other than rulemaking.

The NRC showed no hesitation to provide utilities with the authority to use Part 72.48 to make changes to the design of a generically approved cask so that the cask could be used to accommodate the site specific needs of VSC-24 users.

One might ask, if the utilities which did not design the cask can change the design of the cask without going through rulemaking, why should the vendor of the cask be held to a higher standard of review?

Why 2

shouldn't the vendor also be able to change the design of the cask without going through rulemaking?

It is Citizens' Utility Board's position that the NRC' s decision to allow utilities to use 72.48 was a deliberate attempt to make a generic cask a site specific cask while circumventing site specific license regulations and opportunities for public involvement.

The NRC should rescind the utilities' use of 72.48, and should require all changes to the design of a generic cask to proceed only after rulemaking.

This is ever more apparent since the hydrogen explosion in VSC-24 cask #3 at the Point Beach nuclear power plant on May 28, 1996.

This explosion is being blamed on a chemical reaction that was overlooked by Sierra Nuclear and the NRC during the design and certification of the VSC-24 cask.

Why should we expect that utilities using the 72.48 process to modify the design of the cask to fit their site specific needs will not overlook some element that will lead to a type of accident not previously identified? We can expect that accidents will continue to occur under the current regulations.

The NRC must act to strengthen its regulation of design changes to nuclear waste storage casks, not weaken them any further.

All changes to the design of the VSC-24 cask should require NRC approval following a rulemaking procedure which allows for public participation.

Mrs. Shillinglaw' s Comment Regarding Cask Unloading Procedures Referenced in the NRC' s Supplementary Information Section Appearing in the Federal Register at Vol. 61, No. 94, p 24249 Mrs. Shillinglaw also recommended that the NRC require the disclosure of utilities' unloading procedures for the VSC-24 cask.

Citizens' Utility Board is of the opinion that the generic licensing process for high level nuclear waste storage casks was designed to prevent public involvement in the siting of nuclear waste sites in their communities, as well as to shield the NRC from being held accountable for inevitable failures during the use of the casks.

The NRC should no longer use the language of the Nuclear Waste Policy Act to support the generic licensing of nuclear waste casks.

The use of the VSC-24 cask has illustrated that it is not "practical" to license a generic cask that is modified at each nuclear plant site to meet site specific needs.

Regarding Mrs. Shillinglaw's concern, current regulations allow the users of a generic cask to develop cask loading and unloading procedures which do not require NRC review and/or approval. Because cask users are not required to submit copies of cask loading and unloading procedures to the NRC, the public has no access to these documents through the NRC's PDR.

The well known saga of a defective VSC-24 cask at Palisades remaining loaded with the equivalent long lived radioactivity of 240 Hiroshima sized explosions because of concerns regarding unloading procedures, and 3

the recent alarming explosion at Point Beach during the loading of a VSC-24 cask is sufficient proof that the procedures for loading and unloading of casks should be reviewed and approved of by the NRC prior to use of the cask, and that the procedures should be available for public scrutiny.

In closing, nothing which is stated in the above comments should be construed by the NRC as Citizens' Utility Board requesting a separate petition for rulemaking or as any other petition for NRC action.

Nor should any of the comments be construed to imply that Citizens' Utility Board would consider the VSC-24 dry cask storage system to be an acceptable container for high level radioactive waste if certain regulations were in effect.

Citizens' Utility Board is of the opinion that the VSC-24 is an unacceptable storage system for high level radioactive waste on the shore of Lake Michigan.

The above comments are provided to address Mrs. Shillinglaw' s concerns raised to Chairman Jackson in her December 9th and December 29th, 1995 letters.

Thank you for the opportunity to comment on Mrs. Shillinglaw's concerns.

4

Secretary U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Secretary:

July 15, 1996 DOCKETED USNRC

.96 JUL 30 A 9 :57 OFFICE OF SECRETARY r,'°"\

DOCKET! w & SE~Y er iJv BRA C~

I a~. writing these comments in response to the petition submitted by Ms. t~r:-T.

I Sh1lhnglaw (Docket No. PRM-72-3).

' NU, 1 -R 172_3 Pc f lTION RULC~r.

The past few years of difficulties with the Dry Cask Storage program certainly points to (IJ> \,:f<'2 i4'i changes that need to be made. It is true that in many instances the regulations are written without specifying clear cut steps that need to be followed by utilities and vendors regarding changes after the SAR - SER - CoC are all approved and/or granted. At the time that new generic cask design is proposed, the public has the opportunity to comment on the specific design proposed by the vendor. After the NRG has accepted all proposals, comments and changes, the Coe is granted to the vendor for a generic cask design. For the vendor at a later date to change the design at will makes a sham of the process that the NRG has used (and is mandated by the Code of Federal Regulations).

Clearly, the problems that have resulted at Davis Besse would not have occurred if these regulations had been very clearly defined. The result has been a lawsuit centered on this very issue (which is still pending). The vendor, after stating that the wall thickness of the canisters would be one specific measurement in the SAR (the SER was done on this measurement and the Coe granted accordingly), changed the thickness of the canister with no further input from the public or NRG. After the fact of the change having been made, the vendor then informed the NRG and tried to validate the reasons for change. This essentially cut the public out of the process on this important issue. I believe that Ms. Shillinglaw is seeking specific avenues to be taken if changes are to be made in the SAR.

I0CFR72.48 pertains to a licensee (utility) avenue for making changes in the MRS or ISFSI. This does not state that a vendor may make changes to a cask. Furthermore, this section states that if a licensee desires to make changes that involve unresolved safety questions then licensee must submit an application for amendment of the license, pursuant to 72.56. Again, this section does not state that the vendor may make changes to a cask.

Please take a good look at these regulations and the omissions. These must be an inclusion for the vendor to make changes (if they are to do this), a specific avenue that the vendor must follow prior to making the proposed change, and there must be utility, NRG and public review of the same. The results must be that anytime a utility or a vendor makes a chan e to the SAR that involves an unr viewed safet uestion there must be an amendment and public hearing.

-AUG 1 3 1996 Ackno ed by card...............................,

lJ.S. NUCLEAR REGULATORY COMMISSIO~

DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Pos&mark Date 7f z 2/ -J lo Copes Received_------=,,,_..,,,_.. __

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Additionally, I0CFR72.48 (b)(1)Changes, tests and experiments... details the licensee's responsibility in changes made to the SAR. It states that the licensee is responsible for making the determination about whether a change involves an unreviewed safety question. The licensee is required to have a "written safety evaluation that provides the bases for the determination that the change, test, or experiment does not involve an unreviewed safety question". No where in the Code does it give the same authority to the vendor-making the determination of what is or is not an unreviewed safety question. The Code needs to give authority specifically to who may and may not make changes to a cask design and who may or may not decide what is an unreviewed safety guestion. Also, it certainly seems that the NRC as a regulatory body should not only review and approve every SE but it should also be placed in public domain (PDR) since it is documenting a supposedly proposed change to the original SAR with which the public was involved.

Ms. Shillinglaw is clear in her thinking. The SAR should conform with the SER and CoC. If a vendor wants to propose a change in the SAR, the Code of Fed. Regulations should specifically state the avenue to be followed. I firmly believe that this should be done before the changes are instituted (as was not the case in Davis Besse), should be reviewed and agreed to by the NRC and the utility, and reviewed at a public hearing. This should be done BEFORE changes are made, not after the fact. Again, any change involving a unreviewed safety question should result in an amendment application and resultant public hearing. But if these are to be truly generic casks, then these casks should all be manufactured identically. Otherwise, they must be considered to be site-specific.

The issue of numbering the SARs as revised should be a simple matter to clear up confusion. I agree with that.

Ms. Shillinglaw requests that the unloading procedure be made public by the NRC.

This is not a new request of the NRC. It certainly is a valid request particularly in light of the problems at Palisades with their unloading procedure not be approved by the NRC. The current problems at Point Beach (and perhaps others) further highlight this need. The public has a right to know that these casks can be properly and safely unloaded. We deserve the right to be able to review these documents in detail. I believe that it is irresponsible of the utilities and the NRC to withhold these documents from public review given the current state of affairs with dry cask storage developments.

~~~

3417 Darlington Road Toledo, Ohio 43606 e-mail: CFJohns6@aol.com.

Secretary US Nuclear Regulatory Commission Washington, D.C. 20555 Secretary, 7/16/1996 DOCKET NUMBER ti 7 --°t PETITION RULE PR ~ :

(lol FR 2424~

DOCKETED (0i V

US fJRC ~)

• 96 JUL 22 p 4 :02 OFF/CE nF cr:cp1 T' "'Y I am writing these comments in response to the petition by Ms. F~h~Jllng.,1-aw =~ ?1~r (Docket No. PRM-72-3)

BR A1 Cti

~

In 1990 the NRC, in an attempt to streamline and speed up the establishment of Independent Spent Fuel Installations (ISFSI), inserted into the Gode of Federal Regulations (CFR) provisions for the generic licensing of certified designs of cask for dry cask storage of high level nuclear waste. The generic licensing process avoided site specific environmental impact statements and public hearings as well as the detailed process for change of a cask design detineated in the CFR under site specific requirements. The NRG assured the public that none of the above would be needed because obviously the new "generic" casks were to be suitable for all sites and so no changes of any kind would be needed.

Since the establishment of generic licensing the public has come to realize that there really is no such thing as a "generic" cask. For example, changes to the certified cask designs have been made at Arkansas 1, Point Beach, Palisades, and at Davis Besse.

Some changes have been made to the cask design simply out of negligence and inability to follow the specifications as laid out in the Safety Analysis Reports (SAR).

Vectra Corporation's reduction of the wall thickness of the Nuhoms cask for Davis Besse is a case in point. The CFR for generic licensing contains no avenue for change for these "generic cask" because, of course, they are "generic" and no changes should have to be made. If changes needed to be made, the cask would no longer be generic, but site specific. In my opinion, it is prudent that the NRC recognize and admit that generic licensing is not functional, and that casks for the storage of high level nuclear waste must be licensed on a site specific basis. It is not acceptable that the NRC uses the site specific part of the Code to make changes to a cask which was not supposed to have to be changed, hence "generic".

Certainly, any changes to a SAR for a generic cask (if they must be made) should be done through an amendment process with public comment. Neither the licensee or the vendor should be granted the sole authority to determine if a change involves an unreviewed safety question. The public must be allowed the opportunity to comment upon any proposed change to the SAR or any of the other licensing documents such as the Safety Evaluation Report (SER) or the Certificate of Compliance (COC).

The unloading procedure should certainly be part of the documents made available to the public. Unless the public has the opportunity to scrutinize the unloading procedure, there can be no confidence for those residents near ISFSls that these cask can be unloaded safely.

Comments by Alice Hirt, 615 Islington St., Toledo, Ohio 4361 O JUL 3 O 1996 _

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.J.S. NUCLEAR REGULATOflY COMMISSIOt-.

DOCKETING & SERVICE SECTION OFFICE Of THE SECRETARY OF THE COMMISSION Documerit Statlsb Postmark Date 'J /i q / 9 (p I

I Coples Received._f-j ______ _

Md'I Coples Reprodt.eed _4_..__ __ _

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July 19, 1996 Mr. John C. Hoyle Secretary

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EI DOCK ETED US,RC NUCLEAR ENERG Y INSTITUTE

'96 JUL 22 P 3 :S 7 OFFICE OF S[r".RETA~ Y DOCKET l*'G & :En

. '~E BRA ~CH DOCKET NUMBER PETITION RULE PRU 7 2 -3 Cf)

( hiFR2.424'J)

John F, Schmitt, CHP DIRECTOR RADIOLOGICAL PROTECTION, EMERGENCY PREPAREDNESS

& WASTE REGULATION U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ATTENTION:

Dear Mr. Hoyle:

Docketing and Service Branch Fawn Shillinglaw, Receipt of Petition for Rulemaking (61 Fed. Reg. 24249-May 14, 1996)

Request for Comments The Nuclear Energy Institute (NEI) 1 offers the following comments on behalf of the commercial nuclear industry on the Fawn Shillinglaw, Receipt of Petition for Rulemaking, as noticed in the Federal Register dated May 14, 1996.

Dry storage of spent nuclear fuel is important for the continued operation of nuclear power plants. Increased use of this technology will occur. Challenges which are regulatory and technical in nature could potentially cause significant dry cask storage delays and, in some cases, could impede plant operations.

The petitioner requests that the NRC amend its regulations which govern independent storage of spent fuel in dry cask storage casks to require that a safety analysis report (SAR) for a cask design fully conforms with the associated NRC safety evaluation report (SER) and certificate of compliance prior to the NRC's certification of the cask design. The petitioner also requests that the revision date and identification number of the SAR be specified whenever that report is referenced in documents.

1 The Nuclear Energy Institute (NEI) is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect/engineering firms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry.

. JUL 3 0 199l

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'.:,l_J Tf 40 G WASHII-JCIOI-J DC 20006-] 7()8 PHONc 7()? 739 8000 FAX 202 /85. A019

. J.S. NUCLEAR REGULATOOY COMMISSIO~

DOCKETING & SERVICE SECTION OFFICE Of THE SECRET ARY Of THE COMMISSION ooa,nerrtStatiSb Postmark Date

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Mr. John C. Hoyle July 19, 1996 Page 2 Recently there has been significant regulatory activity regarding the licensing basis for facilities and the status of the SAR and SER in this regard. The practice of maintaining current documentation and the methods for doing so also have been under consideration. These industry and regulatory discussions could influence the merits of this petition and its proper disposition. Enclosed for your information is a copy of NEI's correspondence to Chairman Jackson from Joe F. Colvin, President and Chief Executive Officer, dated July 1, 1996, forwarding an industry evaluation of the regulatory significance of the FSAR and related matters. Pending the outcome of these discussions, response to the subject petition should be held in abeyance.

We appreciate the opportunity to comment on the petition and look forward to discussing this important matter with the NRC after the broader licensing basis issues currently under consideration have been resolved. If you have any questions or need further clarification, please call Alan Nelson at (202) 739-8110 or me at (202) 739 8108.

Sincerely,

~+~

John F. Schmitt JFS/APN/ec Enclosure

Mr. John C. Hoyle July 19, 1996 Page 3 be:

J. Colvin T. Tipton J. Schmitt A. Nelson j:\apn \decom/shillpet.doc

NUCLEAR ENERGY INSTITUTE July 1, 1996 Chairman Shirley A. Jackson U.S. Nuclear Regulatory Commission Mail Stop 0-16 G 15 Washington, DC 20555-0001

Dear Chairman Jackson:

Joe F. Colvin t'*f SIOf. N I ANO CHI EF [Xl"CUTIVf OFFICH The nuclear industry is moving forward aggressively to address a wide scope of NRC concerns related to license basis conformance issues. The NEI Nuclear Strategic Issues Advisory Committee, consisting of each utility chief nuclear officer, has been proactive in addressing these concerns and is considering initiatives for the industry to take to improve the NRC's confidence in each licensee's ability to operate in conformance with its licensing basis.

A necessary adjunct to these activities is to establish a clear understanding of the regulatory framework that establishes the licensing basis to serve as a foundation for future activities. In support of those activities, NEI prepared an evaluation of the regulatory significance of the Final Safety Analysis Report and related matters.

Enclosed is a copy of that analysis, entitled A1wlysis of Fi,wl Safety A,wlysis Report and 10 C.F.R. 50.59 Implementation.

Because this analysis necessarily addresses a number of legal issues, a copy is being sent to the Office of General Counsel under separate cover.

This analysis is intended to serve as a basis for further NRC and industry dialogue on this important topic to begin to address potential differences in understanding and expectations. We will proceed to interact with the NRC staff and OGC on these matters. Please contact me if you have any questions or if you think it would be beneficial to proceed in a different manner.

Sincerely, cZt;c+-~l

~**- Colvin

Chairman Jackson July 1, 1996 Page2 Enclosure c:

Commissioner Kenneth C. Rogers Commissioner Greta J. Dicus James M. Taylor, Executive Director for Operations

ANALYSIS OF FINAL SAFETY ANALYSIS REPORT AND 10 C.F.R. 50.59 IMPLEMENTATION June 1996

I.

Introduction Two issues have arisen concerning licensee conformance to its Final Safety Analysis Report ("FSAR") and the proper use of 10 C.F.R. 50.59 to make changes to the plant as described in the FSAR. During its review of these issues, some members of the NRC staff have stated that, in their view, all commitments and statements in the FSAR should be treated as "stand-alone" requirements and therefore any difference between the FSAR and the plant would be in violation of NRC requirements.

Underlying this position seems to be an assumption that the FSAR is a detailed compilation of a plant's licensing basis.

These developments would impact a long-standing regulatory scheme that is at the heart of the day-to-day operation of nuclear power reactors. There may be specific issues which require further industry and NRC discussion regarding licensee conformance with the FSAR and the application of§ 50.59. However, no evidence has been cited that suggests that the existing regulatory system has not been successful in protecting public health and safety, without requiring conformance with narrow interpretations of each and every statement in a licensee's FSAR, or that the disciplined process imposed by§ 50.59 has been generically misapplied.

This paper analyzes the historical evolution of the FSAR and the NRC's treatment of the licensing basis from a legal perspective and the legal significance of the FSAR under current laws and NRC regulations. In addition, this paper addresses recent NRC staff interpretations of§ 50.59 by outlining the differences between these interpretations and longstanding industry practice. The intent of the paper is to establish a foundation upon which FSAR conformance issues and questions regarding the application of§ 50.59 can be constructively discussed by the NRC and the industry.

As a starting point, it is important to place the FSAR in appropriate context with respect to plant operational decisions. The entire regulatory system is based on the premise that such decisions should be focused on nuclear safety. Because of the complexity of nuclear plants and the comprehensive regulatory system, there must be sufficient flexibility to allow for reasoned decisions in the day-to-day operation of the plant, consistent with a licensee's obligation to comply with NRC regulations and its license, including its technical specifications. NRC regulations define the key generic requirements established to provide reasonable assurance that public health and safety will be adequately protected, and plant technical specifications define the key plant-specific operational parameters that must be met. The FSAR, on the other hand, contains analytical support for these prescriptive operational constraints and general descriptions of the plant and plant systems, structures and components, which provided the bases upon which the NRC first licensed the facility and subsequently as a reference document for analyzing proposed changes.

It has long been recognized by the agency and the industry that compliance with technical specifications and regulations ensure that the plant can be operated safely. The industry also believes that licensees need to comply with FSAR commitments and with § 50.59 requirements. This view, for example, is reflected in the NRC-endorsed NEI Commitment Management Guidelines.1 However, no evidence has been publicly provided that suggests that the well-established enforcement mechanisms already in place to assure FSAR conformance and§ 50.59 compliance are insufficient to address any problems that arise at individual licensees or that otherwise the existing regulatory scheme need be modified or the status of the FSAR redefined. As a result, nonconformances with representations in the FSAR do not necessarily translate into safety issues and do not necessarily mean that an underlying regulation or technical specification has been violated.

II.

FSAR: Background and Historical Perspective In order to understand the legal significance of the FSAR as a licensing document, an understanding of the history of the development of the FSAR is necessary.

Originally, as part of the license application process, applicants prepared what was then called the "hazards summary report." Upon licensing, this hazards summary report became part of the license. In 1962, § 50.59 was promulgated, along with a revised§ 50.36. 27 Fed. Reg. 5491 (June 9, 1962). The two rules together allowed licensees to designate a portion of the hazards summary report as technical specifications. The technical specifications remained part of the license, and Commission approval was required for any licensee proposed change. The hazards summary report was separated from the license, and could be changed by the licensee in accordance with§ 50.59. That portion of the hazards summary report that was not part of the license/technical specifications became the FSAR as it is known today.2 Because of the historical role ofFSARs, they were not written with the precision of regulations, technical specifications or license conditions. Rather, plant systems, structures and components and design analyses were summarily described, and often referred to documents not part of the FSAR for more detailed discussions.

In 1968, § 50.36 was amended to define more clearly what should be included in the technical specifications. The Commission intentionally limited technical specifications to only those matters which require "rigid conditions or limitations on 1

See NEI Memorandum to NEI Administrative Points of Contact re: NEI Guideline for Managing NRC Commitments (January 30, 1996).

2 FSARs were primarily summaries of plant designs and accident analyses. Early FSARs were very compact in comparison with recent FSARs (e.g., two volumes versus today's twenty or more volumes).

2

reactor operation." Portland General Electric Company, et al. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 271-74 (1979). These legally binding controls were derived from analyses contained in the FSAR related to plant features "that are of controlling importance to safety." Id., citing Guide to Content of Technical Specifications for Nuclear Reactors, 33 Fed. Reg. at 18610 (November 1968).

Correspondingly, the FSAR did not have the same legal import as when it had been part of the license, but instead was used to describe the plant systems and analyses that supported, among other things, the technical specifications. Plant changes were controlled from a licensing perspective by§ 50.59, which included a requirement for licensees to maintain records of facility changes made under

§ 50.59 (i.e., changes to the plant or procedures as described in the FSAR). Id.

Before the FSAR update rule (i.e., 10 C.F.R. 50.71(e)), records associated with

§ 50.59 changes were maintained outside of the FSAR. Trojan, 9 NRC at 271-74.

Although§ 50.59 did require licensees to submit to the NRC descriptions of changes, these descriptions were not required to be incorporated into the FSAR.

This meant that until 1980, when§ 50. 71(e) was promulgated, FSARs were not required to be maintained current and consistent with the as-built plant. Thus, prior to 1980, statements and commitments in the FSAR could not have had the status of being stand-alone requirements because the plant would not have conformed to and was not required to conform to the FSAR.

When§ 50.71(e) was promulgated in 1980, the NRC stated that the purpose of having an updated FSAR was "to provide an updated reference document to be used in recurring safety analyses performed by the licensee, the Commission, and other interested parties." 45 Fed. Reg. 30614 (May 9, 1980). The NRC further stated that submittal of updated FSAR pages did not constitute a licensing action. was not intended for the purpose of re-reviewing plants, and was only intended to provide information. In fact, it was stated that the update rule "is only a reporting requirement." Id. at 30615. Section 50.71(e) did require licensees to update the FSAR to accurately reflect all analyses submitted to the NRC with the license application, as part of the licensing review process, as required by § 50.59 and other NRC requirements, or as necessary to support license amendments. Section 50.71(e) did not require any new analyses to be performed. Id.

Further,§ 50.71(e) did not change the scope of the FSAR, although new information was to be incorporated into the FSAR to reflect plant analyses and plant changes that had occurred since the initial FSAR was issued. For early plants, that scope remained narrow; for more recent plants, that scope was already much broader.

From 1980 forward, FSARs associated with new plants grew in size and scope relative to older plant FSARs. largely because NRC requirements were evolving and growing, and additional analyses and detailed plant descriptions were added to the FSAR in response. However, the legal requirement to develop a FSAR (i.e., to support a license application in accordance with§ 50.34(b)) was never changed nor 3

was the legal status of the FSAR ever formally addressed by the NRC. In keeping with§ 50.71(e), the updated FSAR is now used by the NRC to perform various licensing-related reviews and by licensees to document plant information and important licensee commitments. This has ultimately led to greater NRC and licensee dependence on the FSAR as a reference document, and still yet greater inclusion of plant information and details associated with license commitments. In this respect, the FSAR is an important document but, as will be discussed later, that does not elevate the FSAR to a legal status not otherwise established by law or regulation.

In summary, the hazards analysis report/ FSAR, once part of the license, was removed from the license by the NRC in 1962. In accordance with NRC regulations, the plant could be changed from the description in the FSAR under§ 50.59.

Because FSARs were of varying scope and until 1980 were not required to be updated, they were not consistent at any particular moment with the as-built plant, or the same from licensee to licensee. After 1980, FSARs were required to be updated periodically in accordance with§ 50. 71(e) to better support NRC and licensee analyses of licensing actions, but this change did not elevate the FSAR to the legal stature of a license condition. The updating process grew out of a process that involved combining § 50.59 records and other licensing documents into the original FSAR to improve its value as a reference document. The level of detail in these documents varied from licensee to licensee, and over time for each licensee.

For these reasons, updated FSARs were not consistent in their scope or level of detail.

III.

Legal Significance of the Updated FSAR Today The FSAR is an integral part of the license application and, in accordance with

§ 50.57, the NRC must conclude prior to licensing that the facility has been constructed and will be operated in conformance with the application, which includes the FSAR as modified through the time of the NRC's review and issuance of the license. As discussed in detail below, specific provisions in the updated FSAR are legally enforceable over the life of the plant in two respects, that is, to the extent that a provision defines how compliance with NRC requirements will be achieved and to the extent that a licensee fails to change or update the FSAR in accordance with NRC regulations and/or licensee procedures. These enforcement mechanisms offer substantial regulatory control over FSAR conformance. However, the updated FSAR is still not, in-and-of-itself, a legally binding document.

From a procedural perspective (i.e., the mechanism by which a change is made), if a licensee fails to perform a required § 50.59 evaluation, the NRC has treated that failure as a violation of§ 50.59 and has issued Notices of Violation for such failures.

The theory in citing § 50.59 is that a difference between the plant and a related provision in the FSAR is equivalent to a plant "change" that, if not previously 4

evaluated in accordance with§ 50.59, or if inadequately evaluated, is a violation of

§ 50.59. In the NRC's view, this would be true even if that nonconformance were unknown to the licensee and/or had existed since initial plant start-up. The NRC has historically taken this approach and apparently intends to continue this practice. Other potential "procedural" violations for differences between the plant and the FSAR could presumably, depending upon the nature and scope of the nonconformance, be based on 10 C.F.R. Part 50, Appendix B (e.g., Criterion III, Design Control) and§ 50. 71(e) (failure to update the FSAR). In certain circumstances, a violation of§ 50.9, Completeness and Accuracy of Information, could also be cited if the licensee fails to notify the NRC of information required to be complete and accurate in all material respects is not so.

With respect to the substantive aspect of licensee commitments, there is an important hierarchical distinction that has been drawn by the NRC in the past:

commitments in the FSAR are legally binding if a failure to meet a commitment directly linked to an underlying regulatory requirement -- i.e., a regulation, license condition (including technical specifications), or order -- results in a violation of that underlying requirement. Otherwise the failure to meet an FSAR commitment has traditionally resulted only in a Notice ofDeviation3 or, if applicable, a violation of one of the "procedural" requirements described above. Further, the NRC has acknowledged that "a commitment is not an appropriate means to resolve an issue that has a high safety or regulatory significance... Such significant matters are to be included either as conditions of the license or as a part of the plant's technical specifications so that they cannot be changed without the prior approval of the staff." SECY-95-300, Nuclear Energy Institute's Guidance Document, "Guideline for Managing NRG Commitments." This NRC staff position on commitments appropriately makes no distinction between a commitment described in an FSAR or made in any other context.

There is legal precedent establishing the validity of this regulatory approach. See Portland General Electric Company, et al. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 272-74 (1979) (finding that technical specifications represent legal bounds within which the licensee is required to operate the facility); Long Island Lighting Company (Shoreham Nuclear Power Station, Unit 1), ALAB-787, 20 NRC 1097, 1125-26 (1984) (holding that a license condition was not needed to enforce certain licensing commitments because the NRC can enforce commitments in the FSAR by order if necessary); and General Electric Company (Wilmington, North Carolina Facility), DD-86-11, 24 NRC 325 (1986) (finding that enforceable requirements subject to a Notice of Violation are "only requirements specified in 3

NUREG-1600, "General Statement of Policy and Procedure for NRC Enforcement Actions,"

(previously codified as 10 C.F.R. Part 2, Appendix C) provides that Notices of Deviation are "written notices describing a licensee's failure to satisfy a commitment where the commitment involved has not been made a legally binding requirement."Section VI.D., Related Administrative Actions.

5

statutes, NRC regulations, license conditions, or orders;" other licensee commitments are enforceable by agency order or by Notices of Deviation). Further support for this interpretation is found in the NRC Enforcement Policy, which defines a "requirement" as "a legally binding requirement such as a statute, regulation, license condition, technical specification, or order," and, as described above, contrasts that with commitments that have not been made "legally binding requirements."Section IV, n.5 and Section VI.D. Also, the NRC Enforcement Manual, Rev. 1, Chapter 8, pp. 6-9 recognizes an enforcement hierarchy for licensing commitments described in the FSAR.

It appears that recent NRC staff statements run contrary to this established line of legal precedent by suggesting that all statements and commitments in the FSAR should be treated as "stand-alone" requirements such that a difference between the plant and a related provision in the FSAR would subject the licensee to the issuance of a Notice of Violation. Although there is no statutory or regulatory provision that clearly supports this interpretation, it has been suggested that

§ 50.57(a)(2) could be read to elevate the legal status of the FSAR to that of establishing legally binding requirements. 4 The argument is that, because the FSAR is part of the license application, all provisions in it are legally binding in substance (i.e., become "stand-alone" requirements) in the same sense as a license condition. The contrary argument, however, is that such an interpretation would be inconsistent with the NRC's separation of the FSAR from the technical specifications and NRC practice regarding license commitments (e.g., NRC Enforcement Policy).5 In fact, if commitments in the FSAR are deemed to be license conditions, then a licensee could not change such a commitment other than in accordance with§ 50.90, and§ 50.59 would be legally invalid.

Some licensees have a general license condition requiring operation of the facility in accordance with the license application (which includes the FSAR); some licensees have a specific license condition that explicitly mentions the FSAR as the description of the facility being licensed; and other licensees have no such provisions. Depending on the exact wording of the license condition, the FSAR or parts thereof may be alleged to create a "special" legal significance for those licensees. However, it would be illogical to read such a license condition as making the FSAR, per §g_, part of the license.

Further evidence supporting a conclusion that the FSAR, as a stand-alone document, is not "legally binding" can be found in the NRC technical specification 4

Section 50.57(a)(2) states that the NRC must find that "[t]he facility will operate in conformity with the application as amended, the provisions of the Act, and the rules and regulations of the Commission."

5 Such an interpretation of§ 50.57(a)(2) would be "new or different from a previously applicable staff position" and thus would constitute a backfit required to meet the requirements of§ 50.109, Backfitting.

6

improvement program. A fundamental premise of that program is that certain technical specification provisions should be relocated to "licensee-controlled documents such as the FSAR" because (1) those provisions do not rise to the level of significance that necessitates making them part of the license like a technical specification and, as a result, (2) they can be changed without prior NRC approval in accordance with§ 50.59. Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, 58 Fed. Reg. 39132, 39136 (July 22, 1993). The Commission also stated in that Policy Statement that "[c]ompliance with Technical Specifications is required by the Commission, and adherence to commitments contained in licensee-controlled documents is expected." Id. at 39138.

This indicates yet again that the Commission recognized that the licensee-controlled FSAR did not have the same legal standing as the license. To now read all FSAR commitments as legally binding in the same sense as the license would not only undermine the purposes and thrust of the technical specification improvement program, but would also appear to contradict the distinction in the Policy Statement between technical specifications and licensee-controlled commitments.

That having been said, there is no question of what licensees should do. They should ensure that the FSAR is updated to serve as an accurate reference document, that changes being considered by a licensee to the plant or plant procedures should be competently evaluated pursuant to § 50.59, and that any differences between the plant configuration and an associated FSAR description that have safety implications are properly identified.

IV.

Part 54 - Current Licensing Basis It also must be understood that the updated FSAR is not a compilation of the current licensing basis. During the Commission's deliberations supporting the promulgation of 10 C.F.R. Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, the NRC recognized that the current licensing basis

("CLB") was contained in several licensing-related documents, including the FSAR.

The NRC concluded that this approach, in conjunction with existing regulatory and administrative control processes, was sufficient to assure plant safety without requiring a compilation of the CLB into a single, stand-alone document, or a new regulatory process to control changes to the CLB. This conclusion is contrary to a position that the FSAR constitutes a detailed compilation of a plant's licensing basis.

The only regulation that addresses the concept of a CLB is Part 54. In pertinent part, the CLB is defined there as "the set ofNRC requirements applicable to a specific plant and a licensee's written commitments for ensuring compliance with and operation within applicable NRC requirements and the plant-specific design basis (including all modifications and additions to such commitments over the life of 7

the license) that are docketed and in effect.... " 10 C.F.R. § 54.3(a), 60 Fed. Reg. 22461, 22492 (May 8, 1995). The definition further specifies that the CLB "includes the plant-specific design-basis information defined in 10 CFR 50.2 as documented in the most recent final safety analysis report (FSAR) as required by 10 CFR 50. 71 and the licensee's commitments remaining in effect that were made in docketed licensing correspondence such as licensee responses to NRC bulletins, generic letters, and enforcement actions, as well as licensee commitments documented in NRC safety evaluations or licensee event reports." Id. (emphasis added).

The definition of CLB in Part 54 is sweeping in that it includes, for the purposes of license renewal, licensee commitments that are not regulatory requirements and, therefore, can be changed without NRC approval (i.e., licensee commitments made in docketed licensing correspondence such as licensee responses to NRC bulletins, generic letters, and enforcement actions). See Current Licensing Basis for Operating Plants, OPP-92-02, NRC Office of Policy Planning, November 30, 1992, at 5. By its terms, however, the definition of CLB only includes that portion of the FSAR containing plant-specific design information as defined in 10 C.F.R. § 50.2 --

not the FSAR in its entirety. There is no discussion in the supplementary information accompanying promulgation of the final Part 54 regulations and/or the 1995 amendments to the rule which indicates that any portions of the FSAR beyond

§ 50.2 design basis information are considered to be included in the scope of the CLB. See 56 Fed. Reg. 64943 (Dec. 13, 1991); 60 Fed. Reg. 22461 (May 8, 1995).

This, again, is inconsistent with the apparent new NRC staff position that all provisions in the FSAR have the direct force of law.

During the NRC's deliberations on the CLB definition, the NRC clearly acknowledged that the FSAR did not contain all of a licensee's current licensing basis. See e.g., NRC SECY-94-066, Evaluation of Issues Discussed in SECY-92-314, "Current Licensing Basis for Operating Plants" (March 15, 1994). The NRC further acknowledged that those commitments that are outside of the FSAR, but are part of the CLB, are not governed by any NRC regulatory change and notification process, but are controlled by licensee ;:idministrative processes which have been generally effective in managing changes to these commitments and notifying the NRC when appropriate. Id. The NRC found that existing NRC regulatory controls, coupled with licensee administrative processes, were sufficient to ensure that the CLB would be maintained to provide an acceptable level of safety. See Id. at 11-13; see also, 59 Fed. Reg. 46574, 46577, 46582-4 (September 9, 1994) (proposed amendments to Part 54). Further, for license renewal, the NRC specifically rejected imposing a requirement to "compile" the CLB into a single document. See, e.g.,

NRC SECY-94-066; see also, 56 Fed. Reg. 64943, 64952 (December 13, 1991).

Thus, the NRC has previously recognized that the FSAR does not contain all aspects of the CLB, that only portions of the FSAR are part of the CLB and that existing NRC regulatory oversight and licensee administrative processes, even if not consistent across the industry, are adequate to track, maintain, and change 8

commitments to assure safety. This undercuts the notion that the FSAR has been, or should be in the future, a definitive compilation of a plant's CLB.

Correspondingly, an NRC focus on FSAR nonconformances should appropriately encourage licensees to update their FSARs to assure completeness and accuracy, but should not require licensees to compile a detailed recitation of all licensing basis commitments.

V.

Section 50.59 Issues There are two core issues associated with the recent positions articulated by members of the NRC staff regarding the application of§ 50.59, notwithstanding the long-standing practice related to these issues.

The first issue is what constitutes an unreviewed safety question ("USQ") under the criterion related to increases in the probability of an accident or malfunction previously analyzed. Section 50.59 states that an USQ results if the probability or consequences "may be increased." In 1989, the industry completed preparation of NSAC 125, Guidelines for 10 CFR 50.59 Safety Evaluations, to assist the industry in implementing§ 50.59 in an appropriate manner. Simply stated, NSAC-125 provides that small probability increases do not necessarily create a USQ, dependent upon on a safety evaluation of the effect of these increases. Where a change in probability or consequence is so small or the uncertainties in determining whether a change in probability has occurred are such that it cannot be reasonably concluded that the probability has actually changed (i.e., there is no clear trend towards increasing the probability), the change need not be considered an increase in probability. However, a narrow interpretation of this§ 50.59 provision would not allow for any increase, even if the increase is not significant as a matter of nuclear safety. For that reason, the NRC has not endorsed NSAC-125. However, recent NRC guidance6 provides that In considering the acceptability of a licensee's 10 CFR 50.59 evaluation, the staff has found compensating effects such as changes in administrative controls acceptable in offsetting uncertainties and increases in the probability of occurrence or consequences of an accident previously evaluated in the SAR or reductions in a margin of safety, provided the potential increases or reductions in margin are negligible. Normally, the determination of whether there is an increase in the probability of occurrence or consequences of an accident previously evaluated in the SAR or a reduction in a margin of safety and whether such increases are negligible is based upon a qualitative 6

NRC Inspection Manual, Part 9900: 10 CFR Guidance, 10 CFR 50.59, Interim Guidance on the Reauirements Related to Changes to Facilities, Procedures and Tests (or Experiment), issued April 9, 1996.

9

assessment using engineering evaluations consistent with the original SAR analysis assumptions. The compensatory actions must clearly outweigh any potential increase in probability of occurrence or consequences or reduction in margin.

The other major issue relates to safety margins. The NRC Staff has suggested that any facility or procedure change that could result in a decrease in safety margin, as analyzed in the FSAR, even if still within the pertinent regulatory requirement or established acceptance standard (e.g., NUREG-0800), should be considered a USQ.

However, long-standing industry practice has historically not considered small decreases in the margin between the results of a prior analysis, as documented in the FSAR, for example, and the underlying regulatory criteria to be a USQ. Under either approach, the plant as modified still complied with the underlying legal requirements thereby assuring safety, because the result remained more conservative than the regulatory requirement.

When reviewing NSAC-125, the NRC stated that the safety margin "should normally be considered the difference between the regulatory limit (i.e., the limit specified by the regulations or Technical Specifications) and the value of the parameter reviewed and approved by the staff as part of the licensing basis for the plant... [which] is typically the value of the parameter proposed by the licensee in the FSAR as modified by the staff's Safety Evaluation Report(s). This value should be incorporated into the licensee's updated FSAR and is sometimes referred to as the 'acceptance limit.'" The NRC also stated that "[w]here a change in margin is so small or the uncertainties in determining whether a change in margin has occurred are such that it cannot be reasonably concluded that the margin has actually changed (i.e., there is no clear trend toward reducing the margin), the change need not be considered a reduction in margin." Letter from Charles E. Rossi (NRC) to Thomas E. Tipton (NUMARC) dated May 10, 1989.

The NRC's recent positions in the NRC Inspection Manual Part 9900 Interim Guidance on§ 50.59 implementation are inconsistent with the NRC staffs previous position and the resulting industry practice. The Interim Guidance states that "the margin should be measured against that margin described in the facility U[Updated]FSAR, or the staff's SER if the UFSAR does not describe the margin.

This dramatic change in the basis against which a proposed change is to be measured has profound implications.

A question has also been raised regarding the legal significance of NRC Safety Evaluation Reports ("SERs"). NRC SERs have legal significance in that they document the NRC staff's conclusions with respect to whether a facility proposed for licensing meets applicable statutory and regulatory requirements; they establish NRC positions against which the provisions of§ 50.109, Backfitting, can be applied (see NUREG-1409, Backfitting Guidelines, Appendix D, June 1990); and they can reiterate significant licensee commitments (see, e.g., NRC Office Letter No. 34, 10

Utility Commitments, July 311981). With respect to the application of§ 50.59, however, the NRC stated in its review ofNSAC-125 that an NRC SER "is not sufficient to conclude that implementation of the modification does not involve a USQ because such SERs do not normally address the broader implications of a licensee's proposal upon the facility as a whole." See Letter from Jose A. Calvo (NRC) to Warren J. Hall (NUMARC) dated February 26, 1991; see also, Letter from Charles E. Rossi (NRC) to Warren J. Hall (NUMARC) dated December 26, 1991.

However, as noted above, the value of the parameter "reviewed and approved by the staff as part of the licensing basis" is typically the value of the parameter proposed by the licensee in the FSAR as modified by the staffs SER. Thus, the exact legal status of the SER (e.g., in implementation of§ 50.59) is not clear, much like the status of the FSAR subsequent to initial licensing. However, in NSAC-125, the SER serves the role of establishing relevant acceptance standards to serve as benchmarks for USQ determinations.

VI.

The Cumulative Effect Recent NRC staff positions that would alter the existing FSAR / § 50.59 process could also have major unintended effects. For example, a narrow interpretation of

§ 50.59 could require a licensee to seek a license amendment to make a plant change or resolve an FSAR nonconformance that has no impact on safety. Absent Commission relief while such a procedurally focused process is underway, a licensee may be forced to shut the plant down. This could have significant consequences for the public in terms of the reliable supply of electricity and the cost of replacement power, even though the matter has no safety significance. This suggests that a rule of reason should apply in the application of§ 50.59 and the interpretation of the safety and legal significance of statements in the FSAR.

Ultimately, the resolution of what is in the public interest will rest on the NRC policy decision of whether the NRC will continue its historical focus on safety or shift to a narrow interpretation of NRC requirements, without regard to safety significance. This decision will have profound implications to the allocation of both NRC and licensee resources.

VII.

Conclusion Based on the foregoing, an NRC staff position that all license commitments and/or statements in the FSAR should be treated as directly enforceable "stand-alone" requirements and. therefore. any departure from an FSAR provision would constitute a violation of NRC requirements, does not appear to be legally supportable. nor consistent with the agency's long-standing practice, or justified by any change in circumstances. The historical development of the FSAR and the NRC's historic treatment of the FSAR do not support the proposition that the FSAR 11

is a legally binding part of the license upon which enforcement action can be directly based. Moreover, there are well-established enforcement mechanisms already in place to ensure that licensees conform with applicable commitments in the FSAR, (i.e., those "procedural" regulations which control licensee changes to the plant and FSAR), and underlying legal requirements (i.e., regulations, orders, and license conditions, including technical specifications) to the extent that the failure to meet a specific commitment in the FSAR results in a failure to comply with that underlying requirement. There is no need to create a new legal significance for the FSAR.

Additionally, suggestions that the FSAR should have been, or should be in the future, the definitive detailed compilation of a plant's licensing basis, appear to be inconsistent with prior NRC staff assessments of the lack of a demonstrable benefit of compiling a CLB. The staff has already acknowledged that the CLB, spread over several documents including the FSAR, and controlled by existing NRC regulatory and licensee administrative processes, is sufficient to ensure plant safety. No change in circumstances has been identified that would justify a reversal of that conclusion.

Recent FSAR nonconformances found at some plants do not raise safety issues of such significance that the licensing process for the entire industry should be subject to the major upset that will result if the NRC staff continues on its apparent path of modifying in a significant manner, a long-standing regulatory scheme, particularly in the absence of evidence that an economically justified substantial increase in protection of the public health and safety would result (see§ 50.109). Undermining the established regulatory process could have several consequences (e.g.,

unnecessary plant shutdowns) that could have a significant economic impact on the public and the industry with no safety benefit. There is absolutely no question that issues of safety significance have been, and can continue to be, forthrightly addressed under current regulatory processes and practices. There may be specific issues regarding conformance with the FSAR and the proper application of§ 50.59 that need to be addressed. However, there is no need for the regulatory framework to be modified through new interpretations of long-standing requirements and practices.

12

L DOCKET NUMBER P;tib TI/ Of:1 RULE fR~ ~

_3 ti FR 242 Secretary, U.S. Nuclear Regulatory Comm1ss1on Washington, D.C. 20555 Attention:

Docketing and Service Branch DOCKf ED U.S. POSTAL SERVICE>@

US~RC July 16, 199b

<VIA:

Re:

10 CFR Part 72, Docket Number PRM-72-3, Fawn Shi! 1 inglaw, Petition for Rulemaking, Putil 1c Comment

• 96 JUL 22 P 3 56 The petitioner, Fawn Shi l l i ngl aw, requests that NRC a{!}FPftE OF SECRI*,

and clarify agency cegulat ions govecning dry storage DOCKET/ 'G & c:rR*.~IRCr certification.

The petitioner specifically Lequests that BRANCf '{v,E the Safety Analysis Report (SAR) for cask design rul iy conform with NRC Safety Evaluation Report CSER) and Certification of Compliance CCOC) BEFORE (emphasis ctdued>

NRC certification of cask designs.

I also request that NR(

require any change Cs) and/or revisions made lo the S1',K cequire ful I rulemaking amendment procedure with public notification and public comment period.

Fawn Shill inglaw correctly concludes that NPC resolution of dry cask certification as addressed in 10 CFR 72 will have broad implications.

The public is Justifiably concerned about the licensing and performance of all systems tor nuclear waste and spent nuclear fuel management.

Storage and disposal technology, se!Rctlon of specific sites, ~no risk to the general publ le from transportation acc1aent and/or incident are vitally important issues and require stringent NRC oversight to adequately protect the puoi,c health, safety, and welfare.

Communities with existing nuclear facilities should receive assurance that NRC procedure and rulemak1ng µruv1des rur their protection.

Revision of 10 CFR Part 72 to require 3Af for any cask design fully comply with SER ctnd COC beroce certification is granted by the agency is needed 1n NRC rulemaking.

Certification of the VSC-24 cJes19n uy :-.rie procedures requested by the petitioner may have avoided current cask integrity prohlems at Pal isaaes and ~01nt Beach.

NRC certification procedures should mandate that ALL

<emphasis added) agency procedures have been met.

it, dtler licensing, changes in cask design are necessary certifiGation should be suspended until modifications are completed and fully tested, pub! le notification ancl co1mne11t period are provided, and NRC has reviewed ctna appruvea the revised clesign.

Due to the long term risks to human health and the environment from release of the material to be co11t:1.1ried ir1 these casks, I respectfully request that the NRC est~o11sn the most stringent of sL=tndards and fully involve t.t1e µun: JC in agency certification and 1 icensing procedures.

Respectfully submitted, JUL 3 0 l~~b f\cknowi8<1gi<l by canl....... "'.............. ~

a 1

• Ashridge Arnheim Sardinia, Ohio 45171 (513) 446-3135 teleph fax j

l I II I

J.S. NUCLEAR REGUV-.iOOY COMt.~\SSl DOCKETING & SERVICE SEC'TION OFFICE Of THE SECAEi ARY OF THE COMM\SSIOff

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Postmalt Date _.'1..u.,,flliu...

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An. unidentified gas ignit~d in~id~ a nucl~-

, Some of dealt ** with the changes the 1 id ar fuel dry cask at~ the shields and their weld-Point B~ach reactor inb.

rib~iouslv.

all

(.Two ' Creeks~

Wiscon-tti;t; could 'oo wr~~o *was sin)~ shattering safety ndt ant-i'c:ip~ted*~.

claims regarding the There are nihe reac-possible. e:-:plosions in t6rs:"r"\OW- 'Ltsing the cfr,i

_t...,..biliie...ii.. 1.slliiie..,..o...,.f_.rl~c.. x *.,c;.-w

.,~.iait~***""'* --.. __ Gif k Ft?FcUiiiz Mat@il t.J?.g The government and relieve the overcrowd-industry are investiga-ing in their" irr:-adiated tion the ignition ( nu-fue-i *rod *pools.

These clear abolitionists say reactors are:

Po{~t "the e>:plosion")

to Beach~-

Prairie Island~

find the cause of

~he Davi~-Besse~ Palisades.*

unpredictable accident.

Surry~

Robinson~

Oco-Th~ N.RC _issued "Con-nee,~

Fort St.

Vrain ~

firmatqrv Action Let-and Cal~ert Clifts.

ters" to utilities now The NRC

• sent "Con-using cask sto'rage and firmatory Letters" to canceled/postponed some those reactors reaard-olanned fills./

ing the i_gnition.

One "It was an explo-*

letter went to Pali-sion~

because only an sades which recently explosion could movt'J.I ~

• loaded i3 of the sam~

4. 400 pound oiece of type of*. casks~.\One of steel~" said David Mer-wriich is ~leaki~~-

ri tt of the Madison-The

• Lake Michigan based Citizens Review Federati'on has* f.i led a

2;.:-206 with the-Nuclear Board..

• "Something just burning does not move a Regulatory Commission p.iece of steel."

to challenge the*

lack At 2:45 am on 'Mav of

• an

_,adeq1..Late unload-

28.

a welding machine*

ing pr.ocedure., *for the wa~ ignited to weld *the dry cask,~torage system shield lid inside the at Pal i.~ades

• container which had

• .Thes~-casks are what been loaded with rods S~nate ' Bili 1271 and from fhe spent fuel House Resolution 1020 pool.

would

. allow to be The flash/explosion shipped

~11

• over our occurred before the.

hiohwavs until they welding began.

It was fihtf,:;an *11acc:eptable" both heard and observed way of', storage for when it displaced the them.

(The proposed nine-inch thick lid.

acceptable sacrifice An earlier letter-zones for * "temporary" from Marv Sinclair ca storag~ are on Indian major Palisades cask lancb,.)

opponent} to the N.R.C.

oointed out that Wis-consin Electric Power Companv made _over 75 safetv* and handing cha-nges to the VSC-24 CdSk design before.thev ever started loading them.

f.

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U.S. NUCLEAR AE.. lJi.r. TORY COMMISSIO~

DOCKETING & SERVICE SECTION OFFICE OF THE SECRET ARY OF THE COMMISSION Document StatilticS Postmark Date _________

Copes Received. _

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DOCKETED NEVADA NUCLEAR WASTE TASK FORCE, INC(!}~ORATED Alamo Plaza 4550 W. Oakey Blvd.

Suite 111 Las Vegas, NV 89102 702-248-1127 FAX 702-248-1128 800-227-9809 DOCKET MJMBER

/._t;~

" F PETITION RULE PRlt

( I, I F Tl,._ t-t,_ if Cf)

July 11,1996 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 RE: Petition for Rulemaking 10 CFR Part 72

[Docket No. PRM-72-3]

Fawn Shillinglaw

.96 J.l 16 P 3 :54 OFFICE OF SECRETARY OOCKETI G & SERY 1CE BRAHCH Enclosed are comments to the referenced petition for rulemaking.

Also included is one attachment.

The Task Force appreciates the opportunity to comment on this important matter.

-.o..,r~-~J~,

el xecu ive Director

{i) by lJlJL 1 9 l996 gWttt4-.,~ltlti:ttml..Jr~.

H.S. NUCLEAR REGULATORY COMMISSIO~

DOCKETING & SERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION DoaJnent Statistics Postmark Date 2 /r-:, I q 6 Copies Received. __ _.... ___ _

,\dd'l Copies Reprodtad __ 3 _ _ _ _

Special Distr!ootioo Yl+ YJ5; /JO({

L *.LaU< tT, C--z 4. Uc, h-4r

/

Petition for Rulemaking Nuclear Regulatory Commission 10 CFR Part 72

[Docket No. PRM-72-3]

Fawn Shillinglaw The petitioner, Fawn Shillinglaw, is requesting that the NRC amend its regulations governing dry storage cask certification. The petitioner asks that the safety analysis report (SAR) for the cask design fully conform with the NRC's safety evaluation report (SER) and certificate of compliance ( COC) before NRC certification of cask designs. The revision date and number of the SAR would be required to be specified in any references to the report, and the modification process for a SAR must be clarified. The Nevada Nuclear Waste Task Force further requests that NRC require any change or revision made to the SAR be made by way of a full rulemaking amendment procedure with a public comment period.

The petitioner notes in the Federal Register Notice, the example of the faulty VSC-24 cask at Palisades that must now be unloaded. The NRC certified the casks May 7, 1993--the same day that the utility set to begin loading spent fuel into them. (See attachment) The casks were already built and located at the storage site. Even though a workable document for unloading a cask is required by the COC, no such plan was available when the VSC-24 was certified or when casks began to be loaded, or indeed even now.

In July 1994 after cask #4 had been loaded, a defective weld was discovered. As of this date, two years later, there is no approved method for unloading the cask. We understand that the latest proposed plan calls for piercing the shield lid. Had the NRC requirements for dry cask certification been as stringent as proposed by the petitioner, this design would not have been approved and the dilemma with the Palisades VSC-24 cask would likely have been avoided.

Another vivid example of the consequences of the premature certification of the VSC-24 design has very recently occurred at Point Beach. Just prior to welding the cask

lid on a fully loaded cask, an explosion occurred, lifting and tipping the three-ton lid. At this time the cause of the accident and the status of the loaded fuel is unknown. The current situations at both Palisades and Point Beach pose financial problems and, far more important, worker safety problems as remedies are evaluated. These two examples of currently existing situations with the VSC-24 casks clearly substantiate the necessity for the revision of 1 OCFR Part 72 to require that the safety analysis report for any cask design fully comply with the safety evaluation report and certificate of compliance before NRC certification is granted. This would require a tested and approved unloading procedure in the event that unforeseen circumstances arise after the licensed casks have been put in use.

Also, as stated in the petition all confusion concerning revisions to the SAR must be eliminated by requiring correct and complete identification of the revision number and date. Additionally, revisions must be complete and final before certification is issued with revisions subject to a rulemaking amendment procedure including public comment.

At the time of certification, all NRC requirements must have been met. If, after licensing, changes in the design are necessary, the certification must be suspended until any modifications are done, full testing completed, public notification and comment received, and NRC has approved the revised design.

This petition addresses only IOCFR Part 72, but the petitioner correctly concludes that NRC's resolution of it will have broad implications. The licensing and performance of all systems for nuclear waste and spent fuel management, storage, and disposal are of concern to the general public. Communities with operating commercial reactors as well as those being considered for waste sites have serious doubts about the safety of existing and future facilities. The Nevada Nuclear Waste Task Force is submitting these comments because if either the proposed Yucca Mountain repository or a Congressionally mandated interim storage facility become a reality in Nevada, the NRC licensing process will be the ultimate public forum. The NRC license for either or both such facilities will establish the final, irreversible decision regarding public safety and environmental protection.

2

A contractor on the Yucca Mountain project told Nevadans several years ago, in a public meeting, that the NRC is our surrogate - that citizens' anxieties regarding a permanent repository were unjustified because the NRC licensing procedure would guarantee safety. In the case of a repository -- eternal safety. It is therefore not surprising that Nevadans pay close attention to the operation and performance of the current NRC licensed facilities and especia11y to the effectiveness of the licensing process. NRC officials must be aware that the issuance of a license for a cask design under the regulations of 10 CFR Part 72 "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste" has direct implications for all communities under consideration for interim storage sites now or at a future date. Because of the dangers of radioactive contamination, the licensing process must include far broader public input and involvement and be impeccably thorough, complete, and fair.

3

ENVIRONMENT NUCLEAR FUEL RODS ALONG THE SAND DUNES?

A Michigan utility plans to store waste above ground. It's not alone J

ust north of the small town of Covert,.\!ich., Consumers Power Co. officials and envi-ronmental act1\'1sts are locked in a battle that marks a ne\\\\' phase in the nation's long-running struggle o\'er nuclear power. The company's Palisades po\1*er plant reactor needs refueling. But the utilitv has no more room for the spent fuel rods it mu~t_ µlace i11 its water-filled storage pool. So Consumers is taking ad1*antage of a 19Y0 \'ucle:.1r Regulatury Commission rule that lets utiliti""s

~tore waste abo\*e ground with*

out agency re\*ie\\\\*. Pali,ade,; offi-cials plan to transfer older radio-actire fuel rods from its storage pool into concrete and steel silo-like "casks** on a site orerlooking Lake.\1ichigan.

l'nless demon,trators. who ha\*e brandished banners pro-claiming "We Don't \\\\'ant Your Ca,;k(et.~)." finJ ~Olli\:' wa~* to ~top the utilil\*. it \1*ill,;tart mm*ing the fuel into the lli-foot-tall. no-technical difficulties and the fierce oppo-sition of the state's politicians.

Energy Secretary Hazel R. O'Leary, who had to deal wirh the storage short-fall as an executi\*e of \' orthern States Power Co. in.\Iinnesota, has µromised to re\*iew the goYernment's nuclear waste program. But ending the impasse O\'t'r long-term storage does not appear to be a high priority for an Administration preoccupied with other issues. A perma*

nent solution is decades away.

In the meantime, utilities are stuck.

The waste, now about 30,000 metric tons, is projected to rise to nearly 88,000 tons bY 2030 (chart). Almost all the spent fuel sits in water-filled pools designed to cool the radioacti\*e waste. These pools, which were rnpposed to be temporar:-:

,;torage. are running out of room. L'tili-ties such as Consumers Power are "be-tween a rock and a hard place," com-plains company spokesman :r!ark Sa\*age.

CONCRETE SOLUTION. That's \1*hr Con-

,;umers Power opted for the ne\1* de-signs. It cho;;e a ca,:k. manufactured by Pacific Sierra \' uclear Associates of Scotts \'alley. Calif, that recein~J ap*

prO\*al from the '.\RC only on.-~pr. i..-\

spoke;cman for Consumers Po\1*er sa:-*s the concrete-and-steel containers cost some $.')00,000 apiece. compared \1*ith up 1 to S:3 million for competing desigr!s.

Building new storage µoob would be e\'en more expen::;1\*e.

.-\t least two more utilities are plan*

ning to use the Pacific Sierra casks. :-.; k C ton cr,ntainE-rs on.\la\*,. Such OUTRAGED: PROTESTERS AT TH£ PALISADES PLANT Chairman han Selin, a holdo\*er Bush mo\*es are,;ure to be~ome increa,;ingl~*

gan. a gr(Jllp opposed to the ne\\\\' facility.

appointee, defer,ds the Pacific Sierra er,mmon. OYer the next decade. nearly The storage ca,;ks are an int.,rim solu-cask, noting it has undergone extensiH:

half of the nation's 109 operating nucle-tion to a problem that was neYer sup-safety tests and can store high-le1*el ra*

ar plants will run out of space in \1*,1ter*

p(J,;eJ to o.:cur. Construction of a perrna-dioacti\*e waste for up to a century. "We iilled storage pools and be forced to con-nent storage repository for utilities' shouldn't ha1*e to ll'Orry abuut tht:m de-sidl*r abo\*eground storage.

highly radioacti\*e spent fuel was sup-teriorating before a [permant: ntj rt:posi-The Palisades plant i,; tau,;ing a,;tir l"'"**d tu hal'e !,,*en undt-r wa:-* Yt*ars ton* is built." san, Selin.

lil'cau,;e it i;; tht-fir,-t tu t'XJJl*>it the l\J\Hi a,;u But l'n1*ir11nn,t-nt,d L'011l't:l'lb and po-Em-ironmentali,ts and :'l!ichigan offi-

-:kc rule. which dc,esn't r"quirt-utilitie,;

litic*al *ipp,.1sition ha1*,, µre1*e11teJ,1cti,Jn.

cials. who don 't :;hare Sdin's confidence to seek approl':ll for wask*:-t*Jrage,;ik" l'nd,:r current la\1*. th" f"d"ral g,)*;""rn-abc1ut perm,rnem storag,e. ll'ant to block a,; long as the \\\\'a:;te i~ stored in an ment is supposed to take possession of use of the new facilitY-at least until the appron*d container. Before 19:JO. fo*e all spent fuel in 1998 and ha\*e a perma-

-:RC re\*iews the site.. The state attorney other utilities had rtccei\*ed the agency nent storage site ready by :2010. The general's office. which unsuccessfui-O. K. for abo\*e-ground storage-but Energy Dept. has spent more than $3 i....!\,.

ly sought an :-:RC public hf;'aring. is only after a lengthy and exhausti\*e billion studying a permanent considering other legal options.

analvsis of each site.

repo~itory deep under.\ e\*a-p\.

Consumers Po\\\\'er officials need DOO-Mt:D DUNES?.\ow. en\*ironmental-da's Yucca :\fountain. but that some place to put spent fuel when ist;; worry that thtc plants will be using effort has been blo,ked by

~ they recharge the Pabades reactor unpro\*en technolog:-* to store ~--'------------~1--==.. =----

1 -->--1-~--1--~ in June. "\\\\'e're a ~!*mptom of a dangerous nuclear \\\\'aste. \\\\'hat's MOUNTAINS t~**J

~'f_.

larger problem." ~a:-*s Sal'age.

more, Palisadt*s opµon~nts com-The real difficulty rl"ma111,c: find-OF WASTE L.,..

~ -*'t.

plain that the :s;P.(s new proce*

r'9 t-*'*j y

ing a permanent place to put the

<lure doesn't µrol'ide adequate radioacti\*e \\\\'aste. \\\\'hich \\\\'ill be safeguacds. In this me. the High*lml,odiooc>i,o h,, t-j arnu nd foe th ousa,.ds of,ws.

casks will be located 160 \*ards nuclear-plant waSte l'ntil the political system faces from Lake.\!ichigan on

• sand in the U.S. i

t.

up to the challenge, the Pali-dunes \'Ulnerable to erosion.

i sades predicament \\\\'ill be repeat*

"This is just about as sensitil'e 8 = ~°rr°R~~ TOSI

~

ed at nuclear plants across the an area as \'OU could find." de-nation.

clares an outraged ~!ary Sinclair.

_p

• *.'-..,:,r--.... c.--,

By

.lfa,*y Beth Rega n 111 co-chair of Don't \\\\'as.te.\!ichi-

~.'r'~~\(g; l!'a,/1111gto 11 AllMIKIITWION 1991 2000 2010 2020 2030

____ _J B:;Sir-.ESS WEE, M:.:, 10 199] 27 i I; I

- - - - - - - - - - - - - - - - - - - - - - ~

I.IVvl\t: T NUMBS

...f MEMORANDUM TO:

FROM:

SUBJECT:

UNITED STATES 11ETITION RULE.. ~'} J (6 1 FR 1z; 2119)

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 9. 1996 19; t JUL I O P11 3: 0 I ULES David L. Meyer, Chief Rules Review and Directives Branch Division of Freedom of Information Publications Services Office of Administration C. William Reamer C{~v Senior Supervisory Attorney Nuclear Waste Management Office of the General Counsel

.1.,,

~ -

,..u N

I::>

'° N

°'

LETTER FROM FAWN SHILLINGLAW DATED MAY 14, 1996, CONCERNING PRM-72-3 You requested OGC review regarding a May 14, 1996 letter from Fawn Shillinglaw and, in particular, whether it should be treated as a comment on her rulemaking petition (61 FR 24249) or, alternatively, as a correction to the notice of receipt.

Your position is that the letter should be treated as a comment.

We have no legal objection to including the letter as a comment in the docket fi_J e... for the petition.

&J.S. UCLEAR P.t..ii..iUITORY COMMISSIOt-,.

DOCKETING & ERVICE SECTION OFFICE OF THE SECRETARY OF THE COMMISSION DoctJnent StattsticS Postma!( Date _______ _

Cci-s Received _______

.WI Copies Reprod~ed ____ _

~ial Distri tion Jl...&.r:.~t v.LJ.I F co J'\,

DOCKET NUMBER PETITION RULE PRM 1 z-.3 UNITED STATES

(~ I F z. Z.'f-V NUCLEAR REGULATORY COMMISSION DOCKETED Ms. Fawn Shillinglaw 1952 Palisades Drive Appleton, Wisconsin 54915 WASHINGTON, o.c. 20e111 0001 u~l R

-.*r \

March 5, 1996

'96 NAY 22 A11 :08

~ r::.

~,- -:M'.

SUBJECT:

RESPONSE TO LETTERS OF DECEMBER 3, 1995, AND DECEMBER 29, l~:;,

REGARDING THE CONTROL ANO USE OF SAFETY ANALYSIS REPORTS FOR SPENT FUEL DRY STORAGE CASKS LICENSED UNDER 10 CFR PART 72

Dear Ms. Shillinglaw:

In your letters of December 9, 1995, and December 29, 1995, you raise several issues concerning the use and control of safety analysis reports (SARs) for spent fuel dry storage casks licensed under Part 72 of Title 10 of the Code of Federal Regulations (10 CFR).

In the December 9, 1995 letter, you reconnend that the Nuclear Regulatory Comission require that the SAR for a cask design certified under the provisions of 10 CFR Part 72, Subpart L, fully conform with the associated NRC safety evaluation report (SER) and certificate.of compliance (COC) before NRC certification of the cask design. Current practice per-.its cask vendors to confor-. their SARs subsequent to NRC staff issuance of the SER and COC.

Further, you request that whenever the SAR is referenced in docu111ents, the revision date and nUllber be specified.

You note that this has not been done in the past and cite the VSC-24 cask as an example.

You believe that, as a result, there is confusion aaong various parties concerning the appropriate SAR revision to be followed and, hence, a possibility that licensees and/or their agents will fail to adhere to all of the requirements contained in NRC's SER and COC.

You also ask about the procedures to be followed for vendor-initiated changes to the SAR for a particular cask design and suggest that clarification is needed regarding the effect of such changes on general licensees either already using or seeking to use that design.

In the Deceaber 29, 1995 letter, which you ask be considered along with the Deceaber 9, 1995 letter, you repeat the above concerns and describe (and question) changes which you believe were made to the VSC-24 cask by the vendor, Sierra Nuclear Corporation.

As a separate matter, you express concern regarding the unavailability of plant-specific cisk unloading procedures.

Insofar as your c0111ents request that NRC impose requirements on the contents of SARs, they are being considered by the staff as a request for rulemaking as provided for in the Ca.ission's regulations at 10 CFR 2.802.

NRC intends to use the provisions and procedures of Section 2.802 to further evaluate your request and the need for NRC to propose revisions to 10 CFR Part 72.

u._,

.,,-. f:-: LA. TORY COMMISSIUr.

LO:..," '"T/r 1 ~ SERVICE SECTION C,F'CE OF THE SECRETARY ur-THE COMM/ SION Document Statistics Postmark Date.

Cop,~s Received_ I.

Add'/ Copies ReprOduced _..3 tnbutton Pt~-~~.h

---

r -)-~

F. Shillinglaw

- l -

Regarding your question about the requirements for making changes to the SAR fo 11,owi ng issuance of the COC, certain specified changes i nvo 1 vi ng the SAR are allowed under the provisions of 10 CFR 72.48, "Changes tests, and experiments,w or similar provisions that may be contained in the applicable COC.

Changes made under 10 CFR 72.48 (or the applicable COC) that do not require prior NRC review and approval may subsequently be inspected by NRC staff.

As for your questions about the wsummary of WEPCO Changes to VSC-24," the staff has reviewed the contents of the table and found that it lists changes to cask fabrication details.

In this regard, NRC considers Wisconsin Electric Power Corporation's review and approval of the changes proposed by Sierra Nuclear to be consistent with the requirements of 10 CFR 72.48.

Therefore, the changes are not required to be submitted to the NRC for review; however, they are subject to inspection by NRC staff at any time.

In answer to your concern regarding access to cask unloading procedures, plant specific procedures often expand upon the generic procedures described in the SAR and include site-specific information. These site-specific cask unloading procedures are written, evaluated, and retained by, the licensee. Although subject to NRC inspection, the procedures are not generally submitted to the Agency.

As such, they are not available to the public through the NRC.

Sincerely, Wil~~ Ivers, Dfrector Spent Fuel Project Office Office of Nuclear Material Safety and Safeguards Docket Nos.:

50-255, 50-266/301, 50-313/368, 72-5, 72-7, 72-13, & 72-1007 Distr1but1on:

JTaylor*

HTholllpson DMorris, OE@' i CPoland

• ... ~....

LTipton JHannon, NRR GTharpe EDO 0000880/EDO 0000945 JM11hoan JBlaha EDO r/f NMSS r/f SFPO r/f WRussell, NRR NMSS DIR. R/F A~d=

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D rt N.alC S:\nQJ\aD.\GaN'TDL.LTa

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OIC DAU OfflCIAL UCORD COPY

MEMORANDUM TO:

FR<>>I:

SUBJECT:

Jack R. Goldberg, Deputy Assistant General Counsel for Enforceaent Office of the General Counsel Original signed by Gail H. Marcus, Director Elinor G. Adensam for Project Directorate 111-3 G. H. Marcus Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation REVIEW OF A POTENTIAL 2.206 PETITION The attached letter froa Ms. Fawn Sh1111nglaw to Chai~n Jackson, dated Deceaber 9, 1995 (Attachllent 1, *£DO GREEN TICKET 1880*), raises a nUllber of issues regarding the control and use of SAils for spent fuel dry storage casks certified and licensed under Part 72 of Title 10 of the Code of Federal Regulations (CFR).

We understand th~* thts letter was discussed at the Executive Teu Meeting on Friday, January 19, 1996, and that Janice Moore of OGC raised the possibility that the letter be treated as a 2.206 petition, or perhaps a petition for rulemaking.

In order to determine how best to proceed, we request that OGC review the letter and provide rec011111endations in this regard.

We have also attached a si ilar letter dated December 29, 1995 (Attachment 2,

• EDO GREEN TICKET 1945*), in which Ms. Shillinglaw raises additional issues (both procedural and technical) regarding vendor-initiated design changes for certified cask designs.

We request that OGC consider this letter along with the Deceaber 9, 1995, letter in developing its reconnendations.

The EDO's office has requested that we identify how these letters will be treated and propose a schedule for r~spondtng to them by January 29, 199&.

In order to eet this request, we ask that OGC provide us 1ts views no later than COi January 21, 1991.

Thank you for your attention to this utter. Should you have any questions, please contact Andy Kugler on 415-2828.

Att1Chllents:

1.

EDO Green Ticket 1880

2.

EDO Green Ticket 1945 CONTACT:

Andy Kugler 415-2828 DISTRIBUTION:

Central Files CHaughney, fllSS JRoe P033 R/f PEng, NMSS EAdensu Docket No. 72-1007 (w/inc011ing)

OOClllENT NAME:

G:\DRY-CASK\SHIL2206.KAC MME DATE

'-t>P. IIIGc,1 _

~c,

{¥,0*1 I

~

~

.,o OFFICE OF THE GENERAL COUNSEL MEMORANDllt TO:

FROM:

fa;-

SUBJECT:

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20111 0001 January 29, 1996 Gail H. Marcus, Di~ector Project Directorate 111-3 Division of Reactor Projects Ill/IV 0 f e of Nui~a~or Regulation ack. Goldberg, Deputy Assistant General Counsel for Enforcement Office of the General Counsel REVIEW OF A POTENTIAL 2.206 PETITION OGC has reviewed the letters from Ms. Fawn Shillinglaw to Chairman Jackson dated December 9 and 29, 1995.

In the December 9 letter, Ms. Shillinglaw states that the NRC should require that the Safety Analysis Report (SAR) be updated and finalized, in order to include reference to any relevant documents and a revision number and date, before a Certificate of Compliance is issued for generic casks.

In support of this request, Ms. Shillinglaw states that the absence of this inforaation 1n the SAR causes confusion because licensees, vendors, fabricato~s and others often refer to only the SAR, as if it is the only relevant document, when there ay be revi.sions which must be included.

She further states that this practice raises the possibility that an SAR version aay be used which, because revisions have not been included, differs fr011 the NRC staff's Safety Evaluation Report and the Certificate of Compliance.

In the December 29 letter, Ms. Shillinglaw repeats these concerns and raises a number of questions regarding changes to an approved VSC-24 design made by the vendor Sierra Nuclear Corporation and the failure of the NRC to provide cask unloading procedures to the public.

In our view, Ms. Shillinglaw's request that the NRC impose requirements on the contents of SAR's is in the nature of a request for rulemaking, as provided by 10 C.F.R. §2.802, rather than a request for action under 10 C.F.R. §2.206, in that it requests IIOdifications to the C011111ission's regulations. Accordingly, that aspect of her letters should be referred to 8111 Reamer, Nuclear Waste Management Staff, OGC.

Ms. Shillinglaw's questions regarding the VSC-24 casks, including the actions of the NRC regarding the approval of any changes to the cask design and the withholding of cask unloading procedures from the public, aay be answered by a letter fr011 the staff, also in coordination with 8111 Reuer.

cc:

W. Reamer, OGC

UNITED STATES NUCLEAR REGULATORY COMMISSION Ms. Fawn Shillinglaw 1952 Palisades Drive Appleton, Wisconsin 54915 WASHINGTON, D.C. 20655-0001

SUBJECT:

RESPONSE TO DECEMBER 9, 1995, LETTER REGARDING TH CONTROL ANO USE OF SAFETY ANALYSIS REPORTS (SARs) FOR SPENT FUE RY STORAGE CASKS LICENSED UNDER 10 CFR PART 72

Dear Ms. Shillinglaw,

In your letter of December 9, 1995, you raised sev al issues concerning the use and control of safety analysis reports {SARs) for spent fuel dry storage casks licensed under Part 72 of Title 10 of the ode of Federal Regulations (10 CFR).

Specifically, you reco11111end that th Nuclear Regulatory Comission

{NRC) require that the SAR for a cask design ertified under the provisions of 10 CFR Part 72, Subpart L, fully conform wi the associated NRC safety evaluation report (SER) and certificate of compliance {COC) before NRC certification of the cask design.

Furth, you request that whenever the SAR is referenced in documents, the revisio date and number be specified.

You note that this has not been done in th past and cite the VSC-24 cask as an example.

You believe that, as a res t, there is confusion among various parties concerning the appropriate ~R revision to be followed and, hence, a possibility that lic~nsees and/or heir agents will fail to adhere to all of the requirements contained in NR s SER and COC.

Finally, you ask about the procedures to be followed for v ndor-initiated changes to the SAR for a particular cask design and su est that clarification is needed regarding the effect of such changes on ge eral licensees either already using or seeking to use that design.

We appreciate your thoug ful comments and suggestions pertaining to the use and control of SARs for cask designs certified under 10 CFR Part 72 and the potential for confusio with regard to the effects of SAR revisions on general licensees using those designs.

We have carefully considered your remarks in light of our experi~ ce thus far in implementing the design certification and general license proNisions of 10 CFR Part 72 and have concluded that no i11111ediate NRC acti~n is warranted.

/

Our experience,has shown that the types of misunderstandings you believe may occur as a res<Jlt of inconsistencies between the SAR, the SER, and the COC are not a significant problem.

We believe that the relationship of these three documents to one another has been made sufficiently clear to licensees and others in 10 CFR Part 72 and each COC.

Through our inspections and safety reviews, we have verified that licensees understand NRC regulations that make the COC binding on them and that require their review of the COC, SER and SAR before cask use.

In addition, we have verified that licensees are aware they must retain in their records documented evaluations which demonstrate that the cask design bases considered in these reports adequately cover the site conditions of their reactor sites.

We have not seen the types of misunderstandings that you describe.

We, therefore, do not share your views that the SAR must be revised to completely conform with the SER and the

Ms. Shil 1 inglaw the SER and the COC before cask certification and that each documented reference to an SAR must necessarily specify the SAR revision number and date.

With regard to the procedures for vendor-initiated SAR Revisions,,Q1fueral licensees using a certified cask design are required to adhere t the COC, and evaluate the SER and SAR as discussed above, except that certa* specified changes to the SAR are allowed under 10 CFR 72.48, "Changes sts, and experiments." or similar provisions that may be contained i the COC.

The SAR referred to in the SER supporting the COC is, of course, e SAR Revision relied upon by the NRC staff in making its safety dete nation.

We, therefore, do not believe that additional procedural idance concerning the treatment of vendor-initiated SAR Revisions is nece ary. Should a need for such guidance become evident in the future, we wil act promptly to provide it. In the meantime, we will continue our ongoi g effort to identify any areas where clarification of NRC's expectation (through the issuance of additional regulatory guidance or rulemaking could improve the regulatory process for NRC's safety review and licensi g of spent fuel storage facilities.

We appreciate your continued interest *n NRC activities to ensure the safety of spent fuel storage facilities and rust that this response appropriately addresses your concerns.

Should y have any questions regarding this response, please contact Kevin C naughton of the NRC staff at (301) 415-3018.

Sincerely, William T. Russell~ Director, Office of Nuclear Reactor Regulation Docket Nos.:

50-255, 50-266/301, 50-313/368, 72-5, 72-7, 72-13, & 72-1007 DISTRIBUTION:

Docket File(w/inc011ing)

EOO #0000880 JBlaha JRoe OPA FSturz,NMSS AHansen PDIII-1R/F(w/1nc011ing)

ACRS JTaylor WRussell/FM1raglia EAdensu(EGAl)

OCA WKropp,RIII PUBLIC JM1lhoan RZ1... rman JHannon NOlson MParker,RIII PEng VTharpe,NMSS NRR Mailroom (EOOI0880 w/incoming)

JRoe HThompson AKugler OGC CMorrh GKalman DOCUMENT NAME:

PALGT880.LTR

• SEE PREVIOUS CONCURRENCE To reoelw

• copy of 1hie dooument. lnclcat.e In the boa: *c*

• Copy without attaohment/encfoeur* *1*

• Copy with ettachment/encloeure "N" =

No copy OFFICE LA: PD II 1-1

• I E IN:PDIII-1 E PM:PDIII-1* l E NMSS*

l E OGC*

I N NAME CJamerson KConnauahton MGainberoni:clllh CHauahnev WReamer DATE 12/29/95 12/29/95 12/29/95 12/29/95 01/04/96 OFFICE D:ORPV" I N AOPR/NRR N

D:NRR l N Tech Ed*

l N I

NAME GMarcus for JRoe RZ i nmerman I r.:l,-;- IIRussel l tltejac DATE 12/29/95 I I In /96 I /96 1/2/96

I EDO Principal Correspondence Control FRON:

Fawn Shillinglaw Appleton, WI DUE: 12/28/95 Chairman Jackson OR SIGNATURE OF:

• • GRN Russell DESC:

CONCERNING STORAGE CASKS DATE: 12/13/95 ASSIGNED TO:

CONTACT:

NRR Russell SPECIAL INSTRUCTIONS OR REMARKS:

NRR ACTION:

ORPW:ROE IRR ROUTINC:

RUSSELL GILLESPIE THADANI ZIMMERMAN CRUTCHFIELD BOHRER

- I EDO COIITROL: 0000880 DOC M': 12/09/95 FIN.AL REPLY:

CJlC RO: 95-1093 ROUTillG:

'l'aylor llllhoen fflcapson Blaha Paperiello AC1\0N

.w *-

PAPER NUMBER:

ACTION OFFICE:

AUTHOR:

AFFILIATION:

ADDRESSEE:

LITTER DATE:

SUBJECT:

ACTION:

DISTRIBUTION:

OFFICE or THE SECRETARY CORRESPONDENCI CONTROL TICKET CRC-95-1093 mo PAWH SHILLDfGLAW WYOMING CHAIRMAN JACKSON Dec 9 95 LOGGING DATE: Dec 13 95 FILE CODE:

THE NEED FOR COMPLETED SAR, UPDATED TO NRC SER REQUIREMENTS Direct Reply CHAIRMAN, CONR ROGERS SPB'CIAL HANDLING: SBCY TO ACK COIISTITUENT:

NOTES:

DATE DUE:

Dec 28 95 SIGNATURE:

DATE SIGNED:

AFFILIATION:

EDO --- 000880

DOCKET NUMBER PETITION RULE ~

- _3.

i \

(t,, I FR ;z 42J/9j NUCLEAR REGULATORY COMMISSION 10 CFR Part 72

[Docket No. PRM-72-3]

Fawn Shillinglaw; Receipt of Petition for Rulemaking AGENCY:

Nuclear Regulatory Commission.

ACTION:

Petition for rulemaking; Notice of receipt.

DOCKETED

[lj§gb~~l-P]

• 96 NAY -8 P 5 : 11

SUMMARY

The Nuclear Regulatory Commission (NRC) has received and requests public comment on a petition for rulemaking filed by Fawn Shillinglaw.

The petition has been docketed by the Commission and has been assigned Docket No.

PRM-72-3.

The petitioner requests that the NRC amend its regulations which govern independent storage of spent nuclear fuel in dry storage casks to require that the safety analysis report for a cask design fully conforms with the associated NRC safety evaluation report and certificate of compliance before NRC certification of the cask design.

The petitioner also requests that the revision date and number of a safety analysis report be specified whenever that report is referenced in documents.

The petitioner believes that her proposal would eliminate confusion among licensees, vendors, fabricators, and others who often refer to only the safety analysis report as the relevant document when there may be revisions that must be included to ensure compliance with the NRC safety evaluation report and certificate of compliance.

The petitioner also believes that the NRC must clarify the process for modification of a safety analysis report after a cask has been certified.

Q,(U-z..°!

1 /99t:.

DATE:

Submit comments by (-75 da~

i 119 publ i G-ati on i Fl the=feder-a-1--

-R-eqi ster).

Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except as

-J.,).

• cGULATORY COMMIS~/U1' L;;,;c; =*r :' G & SERVICE SECTION OFICf OF THE SECRETARY Or= THE COMMISSION Do:ument Statistics Postmark Dato Copi s Received I

---

~-

Add'I Copies Reproduced

.3 Special D str:bu!ion -P lJ;:-:/2-~,--,.,e,--l:-D ~ Y I

---~~

~-........ *-----*

2 to comments received on or before this date.

ADDRESSES:

Submit comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

Attention: Docketing and Service Branch.

Deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:45 am and 4:15 pm on Federal workdays.

For a copy of the petition, write: Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

For information regarding electronic submission of comments, see the language in the "Supplementary Information" section of this notice.

FOR FURTHER INFORMATION CONTACT:

Michael T. Lesar, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

Telephone:

301-415-7163 or Toll Free:

800-368-5642.

SUPPLEMENTARY INFORMATION:

Background

The Nuclear Regulatory Commission received a petition for rulemaking submitted by Fawn Shillinglaw in the form of two letters addressed to Chairman Jackson dated December 9 and December 29, 1995.

A determination by the Office of the General Counsel on March 5, 1996, specified that the issues presented would be treated as a petition for rulemaking.

The petition was docketed as PRM-72-3 on March 14, 1996.

The petitioner requests that the NRC amend its regulations in 10 CFR Part 72 entitled, "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste."

Specifically, the petitioner requests that 10 CFR Part 72 be amended to require that the safety analysis report (SAR) for a spent fuel dry storage cask design fully conforms with the associated NRC safety evaluation report

3

{SER) and certificate of compliance {COC) before NRC certification of the cask design.

The petitioner also requests that 10 CFR Part 72 be amended to require that the revision date and number of an SAR be specified whenever that report is referenced in documents.

The petitioner believes there is confusion among licensees, vendors, fabricators, and others who often refer to only the safety analysis report as if it is the only relevant document when there may be revisions that must be included to prevent discrepancies between versions of the SAR and the NRC SER and COC for a specific cask design.

The petitioner cites the VSC-24 cask, designed by Sierra Nuclear Corporation, as an example where revisions to the SAR occurred after the NRC SER and COC were issued.

The petitioner believes that no procedures are currently in place to permit a cask vendor to make changes to its SAR after issuance of the NRC SER and COC.

The petitioner also believes that this situation creates confusion and the possibility that an SAR version is being used that directly contradicts SER and COC requirements.

The petitioner asks for an explanation of the process that the NRC used for allowing changes to be made by the vendor to the VSC-24 cask after NRC certification, what were those changes, and how this was accomplished without rulemaking.

The petitioner also recommends that the NRC make cask unloading procedures publicly available.

The NRC is soliciting public comment on the petition for rulemaking submitted by Fawn Shillinglaw that requests the changes to the regulations in 10 CFR Part 72 as discussed below.

Discussion of the Petition The petitioner notes that the regulations in 10 CFR Part 72 establish requirements and criteria for the certification of spent fuel dry storage cask

- -------------------~-

4 designs by the NRC. The petitioner is concerned that no process exists in the regulations for a cask vendor to make changes to a generically approved and certified dry storage cask design.

The petitioner cites the VSC-24 cask as an example where NRC certification was issued for a design that was modified after the actual certification took place.

The petitioner notes that NRC certified the design for the VSC-24 cask on May 7, 1993.

The vendor of the VSC-24 cask, Sierra Nuclear Corporation (Sierra), agreed to submit a revision to its SAR (Rev. OA) for this cask in July 1993, about 3 months after NRC certification, because changes were necessary to meet requirements contained in the NRC SER and COC.

The petitioner states that this revision was never completed and cites an NRC letter to Sierra dated November 28, 1994, which indicated that the SAR still needed modification to eliminate contradictions and differences between the VSC-24 cask SAR and the NRC SER and COC.

The petitioner cites a Sierra submittal dated June 5, 1995, as the first instance where a revision (Rev.

OAA) appears with the necessary changes.

The petitioner also cites a letter from NRC to Sierra which states that Revs. 0 and OA insert material into the SAR that NRC asked Sierra to perform.

However, the petitioner believes that the material appears in the licensing record but not in the SAR.

The petitioner indicates that constant references to the SAR exist in various documents but is concerned that the references do not specify the revision number.

The petitioner believes this creates confusion and the possibility that an SAR version is being used that may even contradict or differ from SER and COC requirements.

The petitioner has concluded that a final SAR for a spent fuel dry storage cask design should be accepted which completely fulfills all NRC SER

5 and COC requirements before the cask is certified. The petitioner also believes that the NRC must address how the final vendor SAR can be modified as needed after a cask design is certified. Currently, the only way an SAR can be amended is through rulemaking.

The petitioner has also concluded that the SAR revision number and date should be required whenever that document is referenced to eliminate confusion and prevent a situation where an SAR does not meet NRC SER and COC requirements.

Lastly, the petitioner is concerned that the NRC is withholding cask unloading procedures from the public and recommends that the NRC make these procedures publicly available.

The petitioner cites an example of a faulty dry cask at the Palisades facility where the licensee has been waiting to have a final unloading procedure approved by the NRC.

The petitioner has concluded that dry cask storage issues should be addressed and resolved by the NRC to set the proper precedent for the national nuclear waste disposal program.

Electronic Submission of Comments Comments may be submitted electronically, in either ASCII text or WordPerfect format (version 5.1 or later), by calling the NRC Electronic Bulletin Board (BBS) on FedWorld.

The bulletin board may be accessed using a personal computer, a modem, and one of the commonly available communications software packages, or directly via Internet.

Background documents on this rulemaking are also available for downloading and viewing on the bulletin board.

If using a personal computer and modem, the NRC rulemaking subsystem on FedWorld can be accessed directly by dialing the toll free number (800) 303-9672.

Communication software parameters should be set as follows:

parity to none, data bits to 8, and stop bits to I (N,8,1).

Using ANSI or

\

6 VT-100 terminal emulation, the NRC rulemaking subsystem can then be accessed by selecting the "Rules Menu" option from the "NRC Main Menu."

Users will find the "FedWorld Online User's Guides" particularly helpful.

Many NRC subsystems and data bases also have a "Help/Information Center" option that is tailored to the particular subsystem.

The NRC subsystem on FedWorld can also be accessed by a direct dial phone number for the main FedWorld BBS, (703) 321-3339, or by using Telnet via Internet: fedworld.gov.

If using (703) 321-3339 to contact FedWorld, the NRC subsystem will be accessed from the main FedWorld menu by selecting the "Regulatory, Government Administration and State Systems," then selecting "Regulatory Information Mall."

At that point, a menu will be displayed that has an option "U.S. Nuclear Regulatory Commission" that will take you to the NRC Online main menu.

The NRC Online area also can be accessed directly by typing 11 /go nrc" at a FedWorld command line. If you access NRC from FedWorld's main menu, you may return to FedWorld by selecting the "Return to FedWorld" option from the NRC Online Main Menu.

However, if you access NRC at FedWorld by using NRC's toll-free number, you will have full access to all NRC systems, but you will not have access to the main FedWorld system.

If you contact FedWorld using Telnet, you will see the NRC area and menus, including the Rules Menu.

Although you will be able to download documents and leave messages, you will not be able to write comments or upload files (comments).

If you contact FedWorld using FTP, all files can be accessed and downloaded but uploads are not allowed; all you will see is a list of files without descriptions (normal Gopher look).

An index file listing all files within a subdirectory, with descriptions, is available.

There is a IS-minute time limit for FTP access.

r l

7 Although FedWorld also can be accessed through the World Wide Web, like FTP, that mode only provides access for downloading files and does not display the NRC Rules Menu.

For more information on NRC bulletin boards call Mr. Arthur Davis, Systems Integration and Development Branch, NRC, Washington, DC 20555, telephone {301) 415-5780; e-mail AXD3@nrc.gov.

Dated at Rockville, Maryland, this gt} day of May, 1996.

For the Nuclear Regulatory Commission.

PRM-72-3

Subject:

DOCKET NUMBER PETITION RULE PAM 1 L -3 G1 FR. 2- 'f Z t/9)

~-~rt"'ED.

l, :*,._,t ; ~*:

~

1996 F~.R 14 Ali 8: 20 Corresponcence from Fawn Shillinglaw dated December 9, 1995, and December 29, 1995 Safety Analysis Reports for Spent Fuel Dry Storage Casks Licensed Under 10 CFR Part 72

v,V, \'>

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i Document Description WE Preparer NRCReview Basis ECR95-0240 Modify MfC Lifting Yoke hooks to improve engagement with MfC K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER trunnions.

T. C. Muehlfcld ECR 95-0241 Modify MrC Lifting Yoke nameplate to incorporate as-built weights.

K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER Rev. 0 and 1 T. C. Muchlfeld ECR95-0254 Revise MSB drawings to clarify vent line weld detail, shorten drain line K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER 21lide an,de and clarifv storue sleeve fabrication/weldinJt T. C. Muchlfeld ECR95-0264 Revise VCC drawings to make alignment plate levelness criteria consistent K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER with the levelness criteria for the liner flan1e.

T. C. Muehlfcld ECR 95-0274 Reduce MSB weld shim width to facilitate installation and removal.

K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER T. C. Muchlfcld ECR95-0277 Revise VCC dnwings to allow use or standard material for VCC lid bolts to K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER facilitate fabrication.

T. C. Muchlfeld

• ECR 95-0282 Revise MSB specification to Rfiect additional material certification K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER reouirements from SAR T. C. Muchlfcld

• ECR 95-0212 was cancel:' after further discussion with SNC and NRC personnel. All parties agreed that cxistirtf§J)S'fjfisili,on was adequate..

The following changes were discussed with the PSCW staff on November 2, 1995.

ECR 95-321 Revise MSB Specification to allow material testing by a company other than T. C. Muehlfcld/

10 CFR 72.212(b)

SeePBNPSER the material suoolier.

A. R Bayer ECR 95-32.S Revise MSB drawings for shield lid to slightly increase clearance for drain K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER Rev. 0and l line, adjust structural lid valve opening to eliminate weld interference, and T. C. Muehlfeld add drain line olu1 to Bill of Materials.

ECR 95-331 Revise VCC Specification to include load testing requirements for VCC T. C. Muchlfeld/

10 CFR 72.212(b)

SeePBNPSER liftina luu.

A. R Bayer ECR9.S-333 Revise MSB specification to allow *Id repair of sleeve comers following K. R Anundson/

10 CFR 72.212(b)

SeePBNPSER the bending 111-

~

T. C. Muehlfcld Additionally, Siem Nuclear Corporation made a number or changes to the approved V~C-24 design prior to fabrication of the Point Beach cask components.

These changes incorporai&I unprovements to the design based on ~ence e!ned from the Palisade's and ANO RWi&fJl* Wisconsin Electric evaluated lhcse changes to ensure there were no unrevicwed safety questions and evaluation have been provicied to the NRC for their review.

r

From:

To:

Date:

Subject:

Emile, Carol Gallagher WND1.WNP2.ELJ 5/31/96 10:20am PRM-72-3 PRM-72-3 was noticed in the Federal Register on May 14, 1996. Please send me a copy of any comment letters you may receive.

Thanks, Carol Gallagher

March 14, 1996 NOTE TO:

Emile Julian, SECY FROM:

Michael T. Lesar, RRDB, ADM

SUBJECT:

CORRESPONDENCE FROM FAWN SHILLINGLAW OGC has determined that the December 9, 1995, and December 29, 1995, from Fawn Shillinglaw should be processed as a petition for rulemaking under 10 CFR 2.802.

Please docket as PRM-72-3.

I have also included a packet of material concerning this correspondence as background that should be included as part of the docket file.

Dall ROUTING AND TRANSMmAL SUP July 10, 1996 10: g:*m*. office aymllol, room number, Initials Date lldln& A,ency/Post)

l.

Fmi l &:> J11l ian

2.

a.

IAdlan FIie Note and Retum IADorova1 For Clearance Per Converution IAa Reauntecl For Correction Prapare Reply Circulate For Your Information See Me Comment lnveatlnte Sllnature Caordlnatlon Justify REMARKS Please place the attached in the file for PRM-72-3, the petition for rulemaking filed by Fawn Shillinglaw.

DO NOT UM this form H a RECORD of approvals, concurrences, disposals, clearancH, and similar actions FROM: (Name, or,. symbol, Agency/PO$t)

Room No.-Bld1.

Michael Harrison, A' 8041-102

• U.S. GPO: 1988 -

241-174 Phone No.

415-6865

. OPTIONAL FORM 41 (Rev. 7-76)

PNscrlbecl 11y GIA FPMR (41 CFR) 101-11.ZOI

Dnl ROUTING

• AND TRANSMl'nAL SUP

.v Jul 16 1996 10: g:m*. office aymllol,,oo,n number, Initials Date ldln1, Aceffq/Poat)

L Fmi le Julian

a.

a.

~

Note and Retum I

For Clearance Per Conversation y IA. Raauested ForCorNc:tlon Pr9pare Reolv Cln:ulate For Your Information See Me k:omment lnvestlnte ISllnature Coordination Justify REMARKS Attached is the May 14, 1996, letter fran Fawn Shillinglaw you requested by telep10ne.

Also attached are two other letters from Fawn Shillinglaw dated May 23 and June 4, 1996 that may not have been sent to you, tut should be placed in the PRM file.

DO NOT use this form ea a RECORD of approvals, concurrences, disposals,

• cleerancH, and similar actions FROM: (Namt; ora. symbol,,Cenc,/Post)

Room No.-Bldg.

Michael Harrison 8041-102

• U.S. GPO: 198B -

241-174 Phone No.

415-6865

. OPTIONAL FORM 41 (Rev. 7-76)

'"'9Krtbecl by GIA FPMR (41 CFR) 101-11.ZOI