ML20059K325

Discusses Request for Commission Comments & Advice Re How Rwrb Should Rule on 930930 Petition Received from Wisconsin Citizens Utility Board (Cub)

ML20059K325
Person / Time
Issue date:	10/29/1993
From:	Murley T Office of Nuclear Reactor Regulation
To:	Kleinhans J WISCONSIN, STATE OF
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ML20059K329	List: ML20059K325 ML20059K331
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NUDOCS 9311150334
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/c October 29, 1993 James Kleinhans, Chairman Wisconsin Radioactive Waste Review Board 3B17 Mineral Point Road Madison, Wisconsin 53705-5100

Dear Chairman Kleinhans:

In your letter to the Secretary of the Commission, United States Nuclear Regulatory Comission (NRC), dated October 6,1993, you requested the Comission's coments and advice regarding how the Wisconsin Radioactive Waste Review Board (RWRB) should rule on the petition dated September 30, 1993, it received from the Wisconsin Citizens Utility Board (CUB).

The CUB petition asks the RWRB to determine that Wisconsin statutes give the RWRB authority to:

(1)

" negotiate with the federal Department of Energy (DOE) regarding the schedule for removal of spent nuclear fuel from Wisconsin Electric Power Company's (WEPCO) Point Beach Nuclear Power Plant (" Point Beach" or "the Plant") to the DOE;"

(2)

" negotiate with the Nuclear Regulatory Comission (NPC) and any other federal agency (in any procedural posture) regarding WEPCO's proposal to construct an Independent Spent fuel Storage Installation (ISFSI) at Point Beach."

(3)

" perform any other functions under [ Wisconsin statutes) sec. 36.50, Stats., (such as serve as initial contact agency, advocate, educational coordinator or negotiator) regarding the DOE's and WEPCO's schedule for removal of spent nuclear fuel from the Point Beach Nuclear Power Plant, or regarding WEPCO's proposed construction of the Point Beach ISFSI."

The NRC staff does not believe it is appropriate to express opinions regarding the scope of the RWRB's authority under Wisconsin statutes. However, with regard to ISFSI construction mentioned in issue (2) of the CUB petition, the NRC would consider appropriate arrangements with the State of Wisconsin for observation of or participation in future NRC inspections of spent fuel storage at the Point Beach Nuclear Plant.

The Comission recently completed rulemaking that approved the VSC-24 cask (proposed for use at Point Beach), a cylindrical concrete cask storage system for spent fuel designed by Pacific Sierra Nuclear Associates (PSNA). A copy of the Federal Reaister notice (SB FR 17948, April 7,1993), including an analysis of the public coments, is enclosed. During the 9-month public coment period, an NRC staff member appeared before the RWRB, on September 3, 1992, to discuss NRC's process for licensing ISFSIs at reactor sites and certifying spent fuel storage cask designs.

In addition, the NRC considered letters from over 200 persons and organizations (including the RWRB's letter 10 0 0 3 u, of September 4,1992). The analysis of public comments in the notice evidences the Comission.s extensive consideration of, and constitutes the tu Comission's response to, the issues raised in those coments.

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q'; James Kleinhans, Chairman. -2'- While the NRC has no comments on the merits of the CUB petition per se, any l negotiations that the RWRB may ultimately undertake with the NRC will of

i necessity have to take place within the context of the Commission's

. independent. statutory responsibilities for the' regulation of spent nuclear fuel..However, the Commission is interested in the public's concerns on spent? fuel storage. issues and will continue to cooperate with..the RWRB and other! appropriate state agencies to. resolve matters affecting publ_ic health 'and-safety.. In addition, to keep the NRC: apprised, Nr.- Roland Lickus,' Chief, j State and Government Affairs, NRC Region 3 Office, will attend the public' hearing to be held on November 4,.1993, at which the above issues-will-be discussed. The NRC will continue to keep the public health and safety of' the citizens of' i Wisconsin and all other states a paramount concern in these activities.. Sincerely, origihal Migned by i no::as z. Nurler Thomas E. Murley, Director Office'of Nuclear Reactor Regulation

Enclosure:

Federal Register Notice (58 FR 17948) i DISTRIBUTION See attached page i t See previous concurrence A PDIII-3 PM:PDIII-D:hII-3 Tech Ed AD:RIII Rill-NRusi brook AHansen: JHannpn

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I DISTRIBUTION Docket File 50-266/50-301 (w/ incoming) l 'NRC PDR & Local PDRs (w/ incoming) EDO # 0009447 JTaylor JSniezek- 'HThompson JBlaha TMurley/FMiraglia- -R Bernero, NHSS JScinto, OGC STreby,' OGC JPartlow WRussell DCrutchfield FGillespie-PDIII-3 (w/ incoming) JRoe JZwolinski-JHannon 0GC OSP/FL (2) Secy # CRC-93-0933 NRR Mailroom (EDO #0009447) (w/ incoming) JPartlow's secretary TGibbons SWiens AGody, Jr. (w/ incoming) AHansen MRushbrook EGreenman, RIII KBohrer (12-G-18) -OPA OCA + I e i i b b i-m

MS48 Tederal Register / Vol. 58. No. 65 / Wedeceday. AprG 7.'1993 / Rides and Repdseless s i i.; 'f i h t 1 l J 3 P 1 1 I NUCLEAR NEGULATORY h it CPR Part 72 notMap Atts l j costa - Assucr. Nuclear Regulatory ) comadenka. I i i j .1

Federal Register / Vol. 58. No. 65 / Wednosday. April 7,1993 / Rules and Regulations 17949 Acnon: Final rule. second cask (TN-24) will be covered hundred and Efty-two assemblies are in separately in a subesquent notice. In storage at Virg!nia Powax. 56 assemblies sutsuAny:The Nuclear Regulatory addition, the comment

• d for the are in storage at CPAL,98 assemblies are o

Commission (NRC) is amending its list June 26.1992, p rule on the in storage at Duke Power and 1462 fuel of approved spent fuel storage casks to VSC-24 cask was reopened to provide elements are in storage at Public Service add one spent fuel storage car.k to the caportunity for public comment on the of Colorado: BC&E anticip:tes loading 1.st of approsed casks. This amendment ebditionst inbrmation (January 21 fuel later in 1993.s will allow holders of power reactor 1993: 58 FR 5301). his comment As a result of b growing use of dry opersting bcenses to store spent fuel in oarfod evpired on February 22.1993, storage technology experience. NRC bas this approved cask under a general Eurther NRC rulemaking activities are gained over 25 staff years of experience license. planned for the TN-24 cask which is, in the review and licensing of dry spent Errtenyt cATE: May 7.1993. therefore, not covered in this notice of fuel storage systems. To further support Acontssts: Copies of the environmental final rule. the NRC technical staff, the egency assessment and finding of no sigr.ificant Section 218(a) of the Nuclear Weste draws upon the knowledge and impact are available for inspection and/ Policy Act of 1962 (NWPA) includes the experience of outside scientists and or copying for a fee at the NRC Public following directive: "He Secretary (of engineers recognized as experts within Document Room. 2120 L Street. NW. DOE) shall establish a demonstration their mpactive fields in the (Lewer Level). Washington, DC. Single Program in cooperation with the private performance of the independent safety copies of the environmental assessment sector. for the dry storage of spent analysis of the systems and components and the finding of no signiScant impact nuclear fuel at civilian nuclear power submitted by applicants for dry cask are available from the individuals listed reactor sites, with the objective of limnses or certiScat!an. Reviews of under the next heading below. establishing one or more technologies numerous app!! cations, seeking either FOR FURTHER e#0Rt1ATt0N CONTACT:Mr. that the (Nuclear Regulatory) site-specific 1SFSis, certificates of Cordon E. Gundersen. Office of Nuclear Commission may, by rule, approve for compliance or approval of a topical Regulatory Research. U.S. Nuclear use at the sites of civihn nuclear power report. have been conducted over the reactors without, to the maximum past 7 years. Regulatory Commission. Washington, DC 20555, telephone (301) 492-3803, or extent practicable the need for Section 133 of the NWPA states, m Mr James F. Schneider Office of additional si:e-spacific approvals by the part, that "the Commission shall by NEclear Materiel Safety and Safeguards. Commission." After subsequent DOE rule, establish procedures for the technical evaluations and bened on a the Cesmisame under sectw, approved liczasing of any technology U.S. Nuclear Regulatory Commission, full review of all available data the n 218(s) for Washington, DC 20555. telephone (301) 504-2692* Commission approved dry storage of use a the site s - e y civilian nuclear spent nuclear fuelin a final rule pwer ree'w sis directive was SUPPL. DAD (TARY edORMATiow: published in the Federal Register on ampleme - ,uly 18,1990 (55 FR Back# ""d July 18.1990 (55 FR 29181).He final 29181)I ir blication in the Federal rule established a ne w subpart K within Re8 ster'of a naal rule establishing e 1 The NRC published a notics of 10 CFR part 72 entitled "Ganeral new e6part L within to CFR part 72 proposed rulemaking in the Federal License for Storage of Spent Fuel at entitled " Approval of Spent Fuel Register on June 26,1992 (57 FR 28645). Power Reactor Sites " Storage Casks." As a result of that 1990 The comment period closed on Irrediated reactor fuel has been rulemaking, four dry casks were listed September 9,1992 but was handled under dry conditions since the in $ 72.214 of subpart K as approved by subsequently reopened, u discussed mid.1940's when irradiated fuel the NRC for storage of spent fuel at below. He proposed rule would have examinstions began in hot calls. Light [ cense.eactor sites under a general wer r amended 10 CFR 72.214 to include two water reactor fuel has been examined additional spent fuel storage casks (i e, dry in hot cells sInce approximately The final rule adde one additional the Transnuclear. Inc.. TN-24 cask and 1960. Some of these fuels have been spent fuel storage cask, the VSC-24 the Pacific Sierre Nuclear Associates, stored continuously in hat cells under cask, to the list of approved casks in VSC-24 cask) on the list of approved dry conditions for approximately two $ 72.214. ne cask being approved, the spent fuel storage casks that powet de::ades. Expertones with storage of VSC-24 cask. is discussed in fur'.her reactor licensees may use under the spent fact in dry casks is extensive. (54 detailbelow.In addition, based on provisions of a generallicense. FR 19379 (1990)). Further, as discussed public comments, the Safety Evaluation Subsequent to the expiration of the below the United States has artensive Report (SER) and Certificate of September 9,1992 public comment arperience in the licanting and safe Comphance br the VSC-24 were period, the NRC took steps to operation ofindependent spent fuel modiSed. Each modification is implemet the provision of $ 2.790(c) of storage installations (ISFSFs). At the discussed below as part of the " Analysis its regulations (41 FR 11808 (1976)) that beginning of 1993 Sve site specific of Public Comments" section ci this provides that information submitted to licenses for dry cask storage had been Federal Repseer notice. NRC in a rulemaking proceeding which issued.They are: Virginia Power's Surry Pacific Sierra Nuclear Associates subsequently forms the basis for a final Station, issued July 2.1988 Carolina (PSNA) submitted a " Topical Report on rule will not be withheld from public Power and Light's (CPAL) HB Robinson the Ventilated Storsgo Cask System for disclosure by NRC. Acx ordingly, on Station, issued August 13.1966:Duks Irradiated Fuel" for their VSC-24 usk January 21.1993, additional Power's Oconee Station, issued January in F 1989. (VSC means information, which was previously 29,1990:Public Service of Colorado's "ventilat stcrage cask." Twenty-far categorized as vendor prietary Fort St.Vrain facility, issued November (24) refers to the number of indimiual information, was in the Public 4.1991: and Baltimore Gas and spent fuel assemblies which the MC-24 Document Room PDR) and all Local Electric's (BGLE) Calvert Cliffs Station. Public Document Rooms. He additional issued November 25.1992. All have TA sere a ort ssx:NT.AF/92-.a.- M n information made availabls on the PDR commenced operation and loaded fuel w om.wy.traU.s.F amim W

• related only to the VSCr24 cask. The with the exception of BGLE.Two test.

17950 Federal Regfrder / Vol. 58. No. 65 / Wednesday, April 7,1993 / Rules and Regn!seinna 1 is designed to hold.)he NRC Public L,. their waste elsewhere.or to shut completed its review and inued its down the plant at Palisades-ril In response to the June 28. M2 and -Conarn saw -_ _ $ that the Safety Evaluation Repet (SER)in A{orJanuary 21,1993. Federal Regisaer review Irocess in!Ost become 1991 approving the Topical Report I nfwencing in a site-speciSc license noticea. 232 comroents were received s.nreasonably delayad and without application. PSNA later submitted its from individuals public interest groups, approval for additi: mal storego approved Topical Report in the form of environmental groupa, associations, capacity, b Pall'.ades plant a " Safety Analysis Report for the industry representatives. Congressional ultimately will b) forced to shut Ventilated Storage Cask System"in representatives, and States. Although a down, a mult that would have November 1991 requesting certification number of the cmnments wem remived serious economic consequences for for use under a general license. The anw the nspective September 9.1992 southwestwo Michigan. NEC cenducted additicnal evaluations and February 22.1993 comment closure -De Federal government's failure to and issued a draft Cert:5cate of datas for the two notices. NRC has resolve questions about b Compliance and dnJt SER, dated April omsinrod comments received permanent storage of nuclear wastes 1992,in support of the Notice of including those received aAer the leaves both'the plant and public with Proposed Rulemaking published in the comment closure detes-limited options: additional storage in Federal Register on June 28.1992. As a part of this rulemaking action, pools, additional storage in dry casks B6 sed on further staff review and NRC remived requests for further or plant shutdown. De federal analysis of public comments, with this opportunity to comment and la government has an obligation to final rulemaking. NRC is approving the particular, for NRC to hold a public resolve h issue of pwmanent or VSG-24 cask for use under a general bearing to review the merits of this interim storega. it would be difficult license and is simultaneously issuing a action. One request was from Frank J. to overstate the need for dispatch in final Certificate of Compliance and SER. Kelley. Attorney General of the State of doing so, as hundreds of American The paramount objective of to CFR Michigan, dated Decernber 30.1992, communities will eventually face this part 72 is protecting the public health which requested a public hearing. problem. and safety by providing for the safe Che.irman Selin responded by letter of -Ten years ago, there wu an erroneous confinement of the fuel and prennting January 25,1993, and proposed a assumption that the search for and the degradation of the fuel cladding. transcribed public meeting with the construction of a final resting place The review criteria used by the NRC for Attomey General to discuss the dry for high level waste would be much review and approval of dry cask storego spent fuel cask app oval promss, to twiner than it has been. A under 10 CFR part 72 consider the answer questions, and to provide

• damanstration" m required by following: Siting, design. quality opportunity for intemsted members of Inw was suppose to sve been for assurance, emergency planning, the public to present comments. nat temporary storage. Because of the training, and physical protection of the public meeting was held on February societal and technical obstacles which 23.199'. fmm 9:30 a.m. until 12 noon radioactive waste disposal presents, fuel included in the nyiew of a speciSc s

systern, either for a cartificate of in Lansing. Michigan,no Attomey even a temporary " demonstration" {egerm is likely to have much rempliance or a site-specific lianse, are General, his staff, representatives of b rogra ten implicatims. Temporary the following: Earthquakes. high winds. NRC staff, and approximately one tomados, tomado driven missilas. hundnd interested citizens attended the dry cask storege in Michigan should lightning, and floods. In eddition, me eting.De meeting was transcribed not bece de facto pomamt disposal. applicants must demonstrate to NRC's and the transcript of that meeting. -It is not fair to the public of Miclu,gan satisfaction that their proposed dry cask including questions and comments of system will nsist man-made events the Attorney General and citizans to link Consumus Poww Company s such as explosions, fires and drop or attending and participatingin the enepts to emune es safe s, tony of its nuclear fuel with the insistence tipover emidents.2 meeting,has been considered b the NRC and is included in the anafysis of by others that we shut down Palisades The VSC-24 cask when used in accordance with the conditions comments. Additionalwritten and every other nucl*** P ant in the l specified in its Certi5cate of comments received within five working. centry. nose comments deal with broad Compliance, meets th nquirements of days subsequent to the meeting have to Cnt part 72.His conclusion is also been considered by the NRC and Policy and prog am iseues relating to reached after a detailed evaluation of an includad in the analysis d the storage and disposal of high-level the VSC-24 cask by the NRC as comments below. (See enmment radioectin waste including the documented in the NRC staffs SER-response number 37 fx information on DePartent dEnwgy's npository Thus, use of the VSC-24 cask, as NRC's response to mquod for a W) . However, mmmenters will a summary o relevant information approved by the NRC, proddes A numbu dumts e nlakd adequate protection of the public health to disposal of high-level weeta, use of on many M 6m bmed inm m me and safety and the environment. dry mak n MW in wal. i'8Ponses to comments set out in NPonse umbem 41. 52. St. and H m Holders of power reactw operating or use of the S424 ca specifically the following analysis of comments limnses under 10 CFR part 50 willbe by Cmsums Poww Corpwatia at h Hany of the comment letters permitted to store spent fuel in this cask Pdh Nudoar Gme4 Suth e ntained comunents that wem simitar under a generallicense. A copy of the ExamP es de indda in nature, nees comments have boen l Certificate of Compliance is available for public inspection and copying for a fue -Cnemes Power Company km grouped as appropriate and addrmed at the NRC Publicrw -t Room. years in advance that the day would as single issues. no NRC hee identified 2120 L Street. NW. (Iower Levolk mm when their rpent fuel and responded to 75 separate issues eat Washington, DC. wouki be fullRey should include the significant points reised by planned abeed of time for this day. each commenter. m.. s.,.i s, $ ,m on.na. m a.,. Consumers Power should be required Many commenters discussed tr pn s wa:=.d enan to cn pan 7: to build a new spent fuel pool, etore that were not the sub)ect of this

~ Federal Register / Vsl. 58. No. 65 / Wednesday April y,1993 / Rules and Reenhtions 17951 rulemaking and thus were not ability of the cask to withstand drop sad activities and possible k4d drop events speci5cally addressed the staff as a tipover accidents. and structural and radiological part of this Anal rulem action.

1. Comment. Socne cxmunenters consequences are necessary evaluations Rese comments expreened opposition eIPressed conceen about b operational under 10 CFR 50.59.

to the use of dry cask storup and safety of the VSG-24 cask relating to For example.the utility's specific included suggestions such as the loading the multi-assembly sealed analyses for load handling activities at following: basket (MSB)into the ventilated the Palisades plant illustrate the type of (1) Nuclear plants generating concrete cask (VCC) and retrieving it. mandatory evaluation by b cask user radioactive waste should be shut Panicularly, the commenters contended that NRC requires before b VSC-24 down. that the loading procedure of placing cask can be used under 10 CFR part 72. (2) The production of redioactive the MSB transfer cask (MTC) on top of subpart K. Among obrs, one specific waste should be stopped when the the VCC is precarious and the procedure event analysed is the evaluation of 6 existing spent fuel pool (and off-fer retrieving the MSB from the VCC is drop of a loaded MTC onto b VCC loed. reactor capacity) is full *, n t clearly explained. One commenter with tipover of the MTC onto the load (3) A formal bearing should be indicated that there are unreviewed distribution system in the track alley required at each sito using dry safety issues associated with handling area.This analysis would encompass Q1 'ng yoke. lugs, and transfer vehicle,uipment including the lifting ca the tipaver scenario descrhd above by storsp cuks. 1 (4) The Palisades Nuclear Plant the commenter who quesdoned whether should be shut down. that need further review. Another it would be part of a utility's $ 50.59 (5)ne embrittlement o'f the reactor commenter arked about the training and evaluation. The result of this analysis pressure vessel at Palisades dictates oversight of personnel performing these shows that the MSB would not fail and that the plant be shut down and no activities. Another asked, that if the that, while local yielding of the transfer additionals ont fuel enerated. transfw cask is on top of the VCCin b cask may occur, the transfer cask would (61The use of uclear wer should ue han&g buhg and a seismk not M and M M N bad to the be stop and existing sites event occurs causing tipover. would this pool for recovery of all spent fuel in the dean uP type of event be considered in a $ $0.59 cask. (7)ne use of storage only casks at evaluation?

2. Co,nment. One commenter Palisades is a vioIntion of public Response. Use of the VSC-24 cask questioned whether. if the MTC were trust; and system in:ide the fuel handling building lifted uby the MSB, the weight of the b

(B) A resaarch and development (including use of b MTC to load and loaded B and the MTC would bear program should be conducted on 7,t feve the MSB from b VCC) would on the MSB welds. Another commenter productive uses of spent l'uel and be conducted in accordance with the to questioned whether & MSB liftin on alternative energy sources-CFR part 50 reactor operator's license. rings could su port the weight of tke Finally, many commenters expressed These cask handling operations. MSB and M1 including loading, retrieval and Response.The weight of the MSB and concern over the ability of dry cask training. must be evaluated by the the MTC could be supported by the storage designs to safely store spent fuel. generafhcensee, as required by to CFR MSB structural weld and the rings. The The following responses to these 72.212(b)(4), to ensure that b weld has been analyzed for this comments reflect a small but important procedures are clear and can be situation and was found to meet & portion of NRC's review of heahn. conducted safely.The MTC and MSB design criteria of paragraphs 4.2.1.1 and safety, and environmental aspects of the have been evaluated against b criteria 4.2.1.2 of ANSI N14.6,1986. This VSC-24 cask, to ensure that the cask is for controlling heavy loads found in standard,which is considered designed to prodde protection of the NRC publication NUREG-0612 conservative,is specifically written for pubhc health and safety and (" Control of Heavy Loads at Nuclear speciallifting devims for shipping environment under both normal Power Plants") and American National containers of radioective materials. This conditions and severe, unl4 aly, but Standards Institute (ANSI) N14.6, situetloo oflifting both the MSB and eedible acz:ldent conditions. Dry cask Special Li&g Devices for Shipping MTCwill rfot ocx:ur under normel storego eystems are musive devices. Containers Weighing 10.000 Pounds or operating conditions. However. if it desiped and analyzed to provide More " The lifting yoke assodated with does occur, as discussed abcyre, the shiwading from direct exposure to b MTCis a spedal purpose device wold and the rings can support the radiation, con 5ne the spent fuelin a designed to ANSI N14.6 criteria to weight of the MSB and MTC. safe storage condition. and prevent ensure that b yoke can safely lift b

3. Comment. One commenter noted releases to the environment.They are wet MTC containing the MSB out of the that tiles at the bottom of the VCC could designed to perform these tasks relying nt fuel pool and can safely lift the break when the MSB is lowered onto ou passive heat removal and MIt and MSB to the top of the them.

confinement systems without moving V &rponse.%ere are numerous parts and with minimal rollana on SpeciBc requirements for libg ceramic tiles arranged on the base of the human intervention to safely fulfill bir yokes, cables, and lugs have been VCC which serve as a separator between function for the term of storap.no identiSed in the Certificate of the Det bottom surfeos of the MSB and designs include margins of safety under Compliance and SER and are not the parallel surface of b VCC liner to both normal and acx:ident conditions to unreviewed safety issues. Part 72 prest the possibility oflocalized provide additional assurance of requires that, prior to the use of a cask corrosion. Although these tiles could protection for the public health and under the general license, b licensee break, there is a substantial margin of safety, the common defunse and security determine whether activities related to safety to prevent breakspe. However, if and the environment. storego of spent fuel under the pneral some breakage occurs, the tiles will still Analyses ofPublic h==* license involve any unroviewed safety perforra their function of providing a questions or chany to the facility slight yp between b MSB and b A. A number ofcommentersinised technical specifications, as provided VCX:. Although it is not necessary, the issues relating to ensk handling and the under to CHL 50.59. Leed handling Certincate of Compliance has been

17952 Federal Register / Vol. 58, No. 65 / Wednesday, April 7.1993 / Rules and Regulations revised to include a statement that the temperature is below 0 *F. and that the safety factor of 5 for a tube against operating procedures for handling the temperature in Michigan and Wisconsin buckling. Because of the conservative MSB over the VCC should include the is often below 0 'F. approach in analyzing a single fuel consideration for reducing the Response.The purpose of restricting storage tube rather than the entire hkehhood of fracturing the ceramic tiles VSC-24 cask movement to ambient basket, the NRC believes that a higher temperatures above 0 *F is to prevent safety factor would exist for the basket by impact load.

4. Comment. One comment'er the possibinty of brittle fracture of the assembly. Thus. the NRC is not questioned why the NRC allows an 80 MSB in the event of a d op accident.

departing from previous design and mch ' ft height when a drop of over 18 There is a 50 'F margin of safety because licansing criteria. J inches may cause enough damsge to the MSB matedal maintains ductile

8. Comment. Some commentsrs noted compromise shielding. Another properties at a test temperature of - 50 that the NRC allowed FSNA to use ccmmenter indicated that the operation

'F. lf a situation for return to the fuel Electdc Power Research Institute (EFRI) of moving the VSC-24 cask from the handling building arises while the rvport NP-4830 in their VSC-24 cask heavy haul trailer across a piece of ambient temperature is below 0 'F a SAR, but did not allow vendors of metal " bridge steel" to the stor,ge pad key option would be for the licenses to casks to reference this report in their sounded dangerous. One commenter determine that the actual MSB material SAR's. also stated that if the MSB is not temperature is above 0 'F. In that event Response. He concept set forth in centered inside the VCC. possible movement of the MSB could be EPRI Report No. NP-4830 is to provida damage could occur to the coating of the accomplished safely without concern for consideration of the cask reinforced VCC linst or the ceramic tiles on the for bnttle fracture.The MSB would concrete bearing pad behaving as a pad s bottom of the VCC. most likely be above 0 *F because of the on an elastic foundation. In previous Response.The NRC evaluated a heat produced by the stored spent fuel. structural reviews of cask systems, the possible drop of the cask and has Another option available to a licensee bearing pad has been very established conditions limiting the lift would be not to move the MSB until an conservatively assumed to be infinitely height for the VSC-24 cask. These ambient temperature above 0 'F is rigid.The response of the pad to a conditions include a requirement to reached. dropped or overturned cask has an inspect the cask after any tipover or

6. Comment. Some commenters stated influence on the magnitude of the force drop from a height greater than 18 that a cask tipover accident while the the spent fuel support system and inches, e.nd the prohibition against VSC is on the pad was not considered, confinement envelope must resist.The lifting the VSC-24 cask to a height even though this type of accident was NRC identified various irsues related to greater than 80 inches. Tha purpose of considered for other asks. Some the details of the concept and its the 80 inch lift condition is to ensure commenters also noted that drop application by the appil. ant.

Rather than relyin that the MSB maintains its confinement evaluations of the MSB were performed report, NRC indepc.g rm the EPRIdently capability even in the event of a drop of for only one orientation although the the VSC-24 cask. The MSB has been

• NRC requires multiple drop orientations the stresses exprienced by the MSB designed to meet the American Society for other designs.

during a drop accident. Based on these of Mechanical Engineers (ASMEl Boiler Response. A cask tipover scrident independent calculations. NRC and Pressure Vessel (B&PV) code under was not specifically performed for the confirmed that the design of the MSB Service Level D conditions and a drop VSC-24 cask.However PSNA will provide an ample margin of safety of 80 inches should only result, at most, performed an en sering analysis of during a drop accident. Therefore, NRC in denting of the MSB shell. The cask drops from vertical and concluded that the design of the MSB purpose of the inspection for any drop horizontal positions which nprosent was acceptable and that there was from a height greater than la inches is more severe accidents than a tipover. reasonable assurance that the to ensure that the shielding is not Therefore.NRC concluded it was not confinement integrity will be comprc..nised and that any damage is necessary to perform a tipover analysis. maintained even if the postulated drop immediately identified arid repaired. With respect to drop orientation, the accident does ocrur. On-site transport procedures with MSB was analyzed for both vertical and In order to provide additional auxiliary equipment such as the " bridge horizontal drop orientations. Information on the application of the steel" described in the Safety Analysis

7. Comment. One commenter asserted concept of an elastichearing pad to Report (SAR) have been reviewed and that the design of the MSB is such that spent. fuel casks. the NRC has initiated are considered to be appropriate to the it is susceytible to buckling under a contract to conduct drop tests of casks design, suitable for use and to meet cartain on-normal and accident from heights in the is to 80 inch range.

safety requirements which are not part conditions.The commenter further his should provide test data that of the regulations in to CFJL part 72. Indicated that this is a departure from would be used to assoas the capability Possible damage to the ceramic tiles was previous spent fuel cask design and of the specific computational techniques discussed in the response to Comment licensing criteria which allow no .ontained in EPRI NP-4830 to predict Number 3. Finally, damage to the buckling of the basket structure. the behavior of dropped casks. coating of the VCC liner would not have Response.The NRC believes that this Fellowing this testing.the NRC will safety signiScance because the liner is commenter refers to the fuel basket and consider the issue of the applicability of not a confinement boundary and does not the MSB shell.The MSB basket the EPRI report, including its not contribute significantly to shielding. structure was analyzed and the NRC applicability to a postulated drop of a concluded that buckling woult not be a steel cask on concrete

• i The principal purpose of the VCC liner is to provide an inner form for the safety concern as discussed below. The

9. Comment.The r #. '!e dynamic concrete during fabrication.

criticalload for buckling was calculated load factor (DII) og k MSB was not

5. Comment. One commenter for a single storage tube and cximpared considered nor was it shown to be indicated that if there were a problem to the actualload under a vertical insignificant.

with a VSC-24 cask. It could not b. deceleration of 124 g that would result Response. The effect of a DLF was removed to the fuel handling building from a drop of 80 inches. The results of conaldered and found to be significant because that is not allowed when the the analysis indicate that there is a ne applicant applied a maximum l

Federal E rian- / Vol. 58. No. 65 / Wednesday, April 7,1993 / Rules and Regulations 17953 4 possible DI.F of 2.0 to the everny Certincate of Complianr=, in Section NRC evaluation documented in section decelerations acting on the MSB. As a 1.2.5? Maximum MSB Rasnovable 4.0 of the SER cocaidored temperature result of using a DIE of 2.0 the Surface Contamination"contains extmmes for both hot and cold decelerstions were incree6ed from 62 g specifications for limiting the amount of conditions. Based on this analysis. b to 124 g and 22 g to 44 g respectively, radioactin contamination permitted on NRC concludes no breach of the MSB for the vertical and horizontal the external surface of the MSB. Rees confinement barrier or leakage from the 1 orientations. As noted above in specifications are coceervative, and are MSB will occur. comment response number 8. althou based. In part, on equivalent criteria

14. Comment. Some commenters j

NRC staff did not endores the meth used for the safe transportation of speculated that a catastrophic release of used by the vendor to determine these radioactive material (see 10 CFR radiation may occur fmm a possible loads the NRCindependently 71.87(i)). Hence, complianos with them explosion caused by spontaneously concluded that these design loadings are will ensure that off-site does limits of flammable uranium hydride in the l accs;ttable. the NRC's regulations will be met for presence of oxygen. It is postulated that

10. Comment. One commenter normal and off-normal conditions alike. the temperatum inside the cask will be provided a cairulation of the results of b generallicenses must also use the hot enough to rupture fuel rods which a hypothetica' accident involving a cask in accordance with the reactor will,in turn, cause h presence of VSC-24 cask. b conditions of the operating license and the Certificate of hydrogen to coats uraalum hydride.

hypohtical accident were a cask Compliance. h general licensee is also Response no NRC does not believe tipover while the caak was under responsible for complying with other that ar explosion inside a storage cask maximum internal pressure. N results Commission regulations regarding caused by flammable uranium hydride indicated that the welds of b MSB radioactivity release limits. brefore, in the pmeence of oxygen is cred$le for would be overstressed, potential releases from the MSB when the following reasons. Oxygen gas is not Response.The NRC reviewed this combined with routine rolesses fmm 6 expected to be present because all cesks calculation and based on that myiew. reactor should not exceed dose limits at are designed to have an inert concluded 6 calculation did not state the site boundary. atmosphere. Further, the formation of the consequences of the hypothetical

12. Comment. Commenters indicated uranium hydride is not cmdible due to occident. Most importantly, the size and that cesks placed close to the shore of the lack of a significant source of configuration of the welds assumed in Lake Michigan represent a serious threat hydrogen. Finally, all casks are designed.

the calculation undstated the strength to the environment, especially to 6 so that the internal temperature will not of the welds and their ability to Croat Lakes which han 20 percent of cause the fuel rods to rupture. withstand the hypothetical event. N the world's surNeo fresh water. Therefore, the conditions necessary for strength of these welds, which meet Response. A utility's use of the this scena.io to occur would not exist. ASME Boiler and Pressure Vessel Code VSC-24 for the storage of spent fuelin

15. Comment. N SER states that criteria, has been thoroughly analyzed casks at a reactor site, would not have there is no credible chain of events that by the applicant and the NRC. Although a signiScant impad on the environment. could spread contamination from the a cask tipover was not specifically his finding is supported by the NRC MSB.Only air coolant loss due to perictmed for the VSC-24 cask, a safety and environmental evaluations blockage was considered. Commenters horizontal drop accident more severe for the VSC-24 cask.inrjuding the indicated that the SER should also than a tipover, was analyzed as a applicant's demonstrou an of consider the effect of Dooding of the hot bounding case. nis analysis compliance of the cask with NRC cask end steam losion. A concern demonstrated that, under the conditions requirements, as well as by b 1990 was also exp regarding the of a borizontal drop while the MSB is rulemaking on dry cask storege and the structuralintegrity of the pods which under raaximum internal pressure, the 1964 and 1989 waste confidence may,in the case of Palisades. be built on welds would not be overstressed.

proceedings. While the VSC-24 cask is a sand dune aree that shifts. B. A numberofcommentersimsed being approved for use under a general Response.b SER for the VSC-24 issues relating to releases of license, it can only be used by a licennee cask did consider the effects of flooding radioactivityfrom surface provided the reactor site parameters as well as alr<oolant loss due to contamination and leclagefmm the (e g., average ambient temperature, bl of the vents.N analysis cosis under normaland occident seismic accelerations, flood water sho the release of contamination conditions. velocity, fires and explosions, etc.), are from the exterior surface of the MSB due

11. Comment. Some commenters ennloped by the czak design basis, as to Gooding is possible but the resultant expressed concern that there would be specified in the SAR and SER. Proper contamis ation would not be significant.

a smal? ~1 ease of radioactive use of a certified storage cask at any site Steam es;losions involving water particWes from the MSB. exterior (whehr near Lake Michigan, a river, a contacting scolten metal are not cmdible surface during off-normal conditions boy, or an ocean) with site parameters under dry spent fuel storage conditions. and that the radioactin releases from that are bounded by the cask design, in addition, explosions due to steem storage casks. when combined with would not have a signiScant impact on forming under flooding conditions are other releases fmrn the reactor, would the environment. not considered credible due to the fact exceed dose limits at the reactor site

13. Comment. Some commenters ht if steam were to be formed,it would boundary.

expressed concem that extremos in be released non-violently through the Response.N NRC interprets this temperatures and humidity would cause vents. comment to mean that during off-dry casks to leak. With resped to b comment on normal conditions there is the potential Response. b VSC-24 cask design structural integrity of the pads, the for relseso of radioactin contamination was analyzed for possible effects of certiScote of compliance requires. per from the exterior surface of b MSB. extremes in temperature and humidity. 10 CFR 72.212(b). that written The consequences of any release of Rose analyses showed no leelage will evaluations be performed by the contamination from 6 MSB exterior occur as a result of temperature or licenses prior to cask use to establish surfeca (whether normal or off-normal) bumidity extremes.b thermal that cask storage pads and areas has e is evaluated in b SAR. However, h analysis presented in the SAR and the been designed to adequately support the

17954 Federal Register / Vol. 58. No. 65 / Wednesday, Apnl 7,1993 / Rules and Regulations static load of the stored casts. 72.122(h)(4) which renda." Storage Chapter 14 of the SER and in Secuon Consequendy, the structural integrity of confinement systems must have the 1.3.1 of the Certificate of Compliance, be the pads would have to be evaluated capability for continuous monitoring in required to verify by a temperature and veriSed before the licensee could a manner such that the licensee will be measurement, the cask thermal use the VSC-24 at the Palisades site or able to determine when corrective performance on a dauy basis to identify at any site. action needs to be taken to maintain safe conditions which threaten to approach

16. Comment. A number ofcomments storsRe conditions." and to CFR cuk desip temperature criteria.The related to gaseous releases from dry 72.122(i) and to CFR 72.128(s)(1) which mak user will also be required to storege casks. Commenters asked the require monitoring of systems and conduct a daily visual surveillance of following questions. What happens to components that an important to safety the cask air inlets and outlets as gaseous components of the decoy chain? over anticipated ranges of normal and required by Chapter 14 of b SER and Are they released to the environment? If off-normal operation. Also, one Section 1.3.1 of the Certificate of not,is prwsure buildup over time being commenter suggested that because the Comphance.

considered? A commenter expressed the VSC-24 cask requires surveillance to While the MSB and VCC are opinion that casks should have ensure that the vents are not blocked. considered components important to individual rsdionuclide emission the requirement that the cooling system niety that comprios b VSC-24 cask monitoring. An issue was raised about must be e passive system (10 CFR daign.they are not conddered the effects of release of krypton.85 72.236(f))is violated, operating systems in the same sense as %r-85) gas on electnc conditions in the Response. NRC approval of the VSC-spent fuel pool cooling water systems or atmosphere. 24 cask system is not inczmsistent with Response.The Easeous components of 10 CFR 72.122(b)(4). 72.122(i) or ventilation systems which may require other instrumentation and control the decay chain are orpected to be 72.12B(a)(1). Although the cited sectims systems to ensure proper functioning. retained within the matrix of the spent of 10 CFR part 72, subpart F. refer to Hence, due to this possive design, fuel or within the fuel rod. In the case " monitoring" or " continuous temperature monitoring and of pinhole leaks in the fuel rod monitoring." they do not specify the survMnar ce activitiu are appropdate cladding, the MSB is designed as s details for particular monstonng and sufficient for this design.they secondary confinement barrier to retam, programs to allow the NRC to require assure adequate protection of the public gaseous products. Therefore, because no monitoring programs that are gaseous components are released to the appropriate for the particular storage health and safety, and meet the environment, no routine monitoring of system design. The NRC has and will requirements of 5 72122 (1). effluent from the outlet vents is consider continuous monitoring where

15. Comment. Several commenters requirsd.The primary reason for it believes contmuous monitoring is erPressed concern related to b inlet requiring the use of ASME section Ul needed to determine when corrective and outlet vents. on the VSC-24 cask, instead of other standards is to ensure action needs to be taken. To date, under which are n==ury to anow coonng of th 8'

tain ns the confinement of fission products. 6 ' general license, NRC has accepted Pressure build-up of gsseous continuous pressure mozutoring of b d h components in the MSB is not inert helium atmosphere as an indicator questi,oned the adequacy of the significant due to 6 age of the fuel and of acceptable performa.nca of surveillance requirements for the VSC-integrity of the fuel rod claddinky; r.ed for mechanical closure seals for dry spent 24 cask and suggested that electronic however, the MSB has been ana fuel storage caska. continuous monitoring and recording of a hypothetical condition in which all ne NRC does not consider such air outlet temperature should be the fuel rods rupture. The resulting contin aous monitoring for the VSC-24 required on each cask. SpeciSc concerns pressure within the MSB is negligible. cask double weld seals to be neaaury include: c The purpose of maintaining an inert because:(1) ners are no known long-(s) Vent blockage by bugs, webs, atmosphere in 6 MSB cavity is to term degradation mechanisms which snow, and ice; ensure that fuel rod cladding would cause the seal to fail within the (b) Frequency of vent outlet degradation does not occur, thereby design life of the MSB and (2) the surveillance for blockaE*: preventing gross fuel rod cladding possibility of corrosion has been (c) Drive-by or walk-through rupture. In addition to ensuring that included in the design (See SER Section ins on is inade:ruste to observe new pin hole leais do not develop in 5.3.1). These conditions ensure that the ou Mblockage;and the fuel clad during the storage period, intemal bellum atmosphere will maaln l the licensee is responsible for nerefore, an individual continuous (d) Critical tem atures associated with the VSC be monitored. monitoring the environment within the monitoring devics for each MSB is not i to ensure that neessaary. However, the NRC considere RerPase.The NRCis mquiring, as MSB prior to its openinfredioactive no unplanned release o that other forms of monitoring casks part of the VSC-24 CertiBcate of 1 meterial takes place. The amount of Kr-including periodic survemanm, Compliance, that surveillance and 85 that could be potentially released inspection and survey requirements, measurement of the thermal from d,y cask storage is so small that it and application of preexisting Performanca of the cask be conducted would not signincantly affect the redlological environmental monitoring by the limnaa= on a daily basis.The ph sics or chemistry of the atmosphere. Programs of part 50 licensees during the licensee is responsible for establishing

d. A number ofcomments wer, period of use of the MSB canisters with the speciSc mahod of measurement, the reccind thatfocused on monitoring, seal weld closures can adequately limnaes can measure the inlet and surwillonce, and inspection actinties satisfy the requirements of 10 CFR outlet air annulus temperatures, or it associated with dry cosk sforege of 72.122 (h)(4).

could also measure the MSB surface s(nt fuel porticularly os they relate to With respect to the issue of temperature,the VCE inner wall t VSC-N cost. Instrumentation and control systems to temperature or perform other

17. Comment. Some commenters monitor systems which are important to appropriate measurements. The method suggested that, with respect to the VSC-safety (10 CFR 72.122 (i)), the user of the setected by b licensee must prende a 24 cask, the NRC did not enforce to CFR VSC-24 cask will, as provided in positive indication of 6 approads of

Fedaral Register / Vol. 58. No. 65 / Wednesday. April 7.1993 / Rules and Regulations 17955 materials ta cask dwign temperature resuha indicate that h shear and lock of centering of 6 MSB inside the criteria. bearing capecities of the concete VOC. nerefore. no precia centerin In addiuon. analyus of safety margins surmunding the airinlet vents (per b MSB inside the VCCis needed. g of of components important to safety show American Concrete Institute (AO) However, the physical arrangement of that even assuming surveillance were criteria 349-45) are not exceeded and no the system restricts lateral movement not conducted at tus required daily damage is expected. nerefore, there is and does not allow the MSB to be far frequency, and both the inlet a.nd outlet no need to insped vents for damage imm center as it is lowered into the vents were blocked for a 30 hour period, foHowing use of the hydraulic roHer VCC. 4 there would sull be no loss of safety skid.

24. Comment. One commenter raised function or any immediate threat to the

21. Comment ne generallicensee the concem that the VCC concate heshh and safety of the public.nis must have spec!Sc plans for b temperatures do not comply with the conclusion is based on the adiabatic constant and careful monitoring of the ACI-349 temperature criteria.

1 hutup thermal analysis of the VSC-24 casks and for 6 safeguarding of b Response. The NRC has accepted ) cask, which assumes that all vents are waste to prevent catastrophic accidents deviations from b AG-349 Code, i blocked. and no heat is re}ected by the or terrorism. Appendix A.4 for the concrete cask. De concrete and cladding Response. In accordance with 10 CFR temperature criteria. However, while temperature criteria that could be 72.212(b)(5), each reactor licenme must accepting the deviation, the NRC has 1 exceeded under this conservative have a physical security organization idactified a speciSed maximum thermal analysis, assuming complete blockage, and program to detect intrusion into b expansion coefEcient for fine and coam signify the onset of very slow pmtected area includir.g acts of eggregates in the concrete which allows degradation mechanisma, not an terrorism, and to take any corrective operation at higher temperatures. The imminent loss of safety function. action. De physical security program, selection of speciSc Ene and criarse ne NRC also agrees with the as well as environmental monitoring aggregetes in the concrete prevents ccmment that visual surveillance of and radiation protection programs for micmcracking between the cement and exterior air inlets and outlets may be each reactor facility, provide the aggregates in the anticipated inadequate and may not lead to e necessary monitoring for b casks and temperature raage of b VCC. Thus, positive determination of blockage safeguarding of the spent fuel. Thus, the deviation from the AQ-349 temperature because the design includes screens licensee will be able to determine when criteria is not a cause for concern and placed over the vents to prevent wildlife corrective action needs to be taken to does not compromise safety. from entering the VCC. Consequently, maintain safe storage conditions to

25. Comment. One commenter the NRC has revised the CertiScate of protect the public health and safety, claimed that NRC.has used the Compliance surveillana requirement to (Also see respons to Comment Number unsupponed assumption that 48 hours make 6 integrity of 6 screens be 33 below).

is sufEcient time to reach thermal of the visual surveillance. A physi D. A number ofcommenters rnised equilibrium for the irradiated fuel examination of the vent is required ifits technicalissues related to the thermal assemblies (high level ndioactive associsted screen shows any evidence of analysis of the VSC-Jd cask and waste) that have been removed from breach. thermalperformance of the cask under water storage and sealed in the rnetal

19. Comment. One commenter normal, off. normal, and occident canister.

suggested ht approval of the VSC-24 conditions. Response.no commenter refers to cask should be denied because the snow

22. Comment. One commenter the time period allowed for a loaded shield was eliminated and that the questioned whether NRCintends to VSC-24 cask system to reach thermal analysis of air flow of the VSC took it establish 75 'F as a standard ambient equilibrium conditions. For the purpose into consideration.

temperature criteria for all storage casks of thermal equilibrium, the VSC-24 cask Response.The snow shield was and expressed concern that this system is cxansidered to be placed in eliminated because it was not tempersture may not be app!! cable for servicz when the concrete cask cover cc.usidered effective in resolving b b majority of power reactor sites. plate is installed. proMem of vent blockage by snow. A Response. The NRC does not intend to It should be noted that the Certi5cate vimal surveillance requirement is establish 75 'F, or other standard of Compliance has been changed to considered more effective in addresung ambient condition criterion, for all cask require that the inlet and outlet air the inuo of vent blockage by snow. The designa. The cask vendor establishes temperatums, for all VSCs placed in CertiScate of Compliance has been ambient temperature criteria on which service,be measured until the cask revised to add a daily surveillance the cask is designed. In the case of the reaches initial thermal equilibrium. i requirement, as discussed in Comment VSC-24 cask, PSNA chose 75 'F. Each Furthermore, a daily measurement of 28, which would include checking for reactor licensee con then only use those the thermal performance of the VSC-24 snow blockage during periods of snow casks which have design bases that cask is required.Thrrefore, any accumulation. In adnition the inclusion envelop the reactor site ambient reforence to assumed 48 hour thermal of a snow shield in the original design temperatures. For exampla,if a power equilibrium is covered by the enhanced actually decreased air flow and reactor site has an average annual surveillance requirements. The 48 hour there fore, its removal increases the ambient temperature greater than 75 'F, period was selected to provide a basis thermal efficiency of the cask. then that reector licensee cannot une a for baseline inessurements. There is no

20. Comment. One commenter cask with a 75 'F ambient design safety signiacanca if thermal questioned how the condition of the temperature.

equilibrium is achieved in a shorter or inlet vents is checked for damage after

23. Comment One comm=ter longer time.

the lifting arms are inserted into 6 air questioned how best transfer for the

26. Comment. One commenter noted inlets for transfer.

VSC-24 cask is affected b he fact that that in chapter 9 of the SER, the NFC t Response. Lifting the VSG-24 cask there are no provisions La centering the staff found it necessary to impose a pro-using the hydraulic roller skid, which MSB inside the VOC. operational test to verify the heet involves insertion oflifting arms into Response. Heat transfer for the VSG-removal capacity of the VSC-2s cask the air in!sts, has been analyzed. N 24 cask is not signi$cantly affected by system. ne commenter claimed that i

17956 Fedemi Kegneter / Vol. S8, No. 65 / Wednesday, AprG 7.1995 / Rules and Regulations this was required because predicted fuel actual temperetun will be lower than temperature testmg of b c sk with a 24 clad temperatures are a " mere" 4 'F the calculated tempereture, canaidering kW loading, the licanaes would be able below their design crHerie ce a 75 'F uncertainties, and therefore this 4 T to lood fuel et lower th utnal stings ambient day. It was further asserted that margin below the fuel clad tempe eture without the need to pro -ide NRC with with a predicted fuel clad temperature criterion is acceptable, separate tempereture test i.;e and of 4 Y telow design cnteria for the off.

27. Cominent. One commanter additional snalysis since the 24 kW host normal condition hmit, even a questioned whether cladding failures loeding is a bounding analysis.

successful pre. operational test would would a!Tect the temperature of the MSB However, becoues the cask vendor has not assure that the design criteria is met or the VCC and the heat removal not provided thermal analyses at lower withm the bounds of statistical capacity of b VSC 24 cask. Another heat loadings, the NRC blieves that if uncertainty, particularly since the asked why helium was used to fill the a licenWs first fuelloading has a heat calihrstion of their temperature sensing cask. N only helium cooled reactor in load less than 24 kW the licensee equipment has a tolerance of plus or the country. Ft. St. Vrain, was should condud both a temperature minus 1 T. operational memly 15% of the time. measurement and a thermal analysia. Response ne NRC has impceed a Response. Fuel cladding failure is not The purpoos of conducting both the test to benchmark the best removal expected to ocrur because the VSC-24 analysis and the measurement is to especity for the first VSC-24 cask czak is designed to maintain sa inert messun system performance and to plead in-service. However, the 4 T helium atmosphere inside the MSB to establish baseline data for the expected margin stated on page 9-e of chapter 9 prevent fuel cledding failure. However, inlet and outlet temperature diffenncs. of the SER cited by the commenter, is ruel cladding failure would neither b CertiScate of Compliance has been a typog sphical error.The corred affect the tempereture of the MSB or revised to this effect e.nd the word trargin is 24 'F, as stated on page 4-7 VCC nor affect the heet removal

• sppmximately" has been deleted. With of the SER. This 24 T margin is the cmpacity of the VSC-24 cask.De respect to the issue of artificial thermal differenca between the maximum temperature of the MSB and the VCX1 loads, the NRC will acx ept alternate heat e:lowable fuel clad tempereture and the depends on the hest genersted by the loads other than spent fuel and the calculated fuel clad temperature, fuelin the MSB, which is not affected Cert 15cate of Compliance has been assuming an averspo annual ambient by a fuel cladding failure. In addition, revised acx:ordinFJy. A licensee could temperature of 75 T for no mal heet removal especity of the VSC-24 use such an artificial heat source to test continuous conditions. For off-normal cask depends on the airflow on the an initial caak at a bounding heat load conditions involving higher ambient outside of the MSB which also is of 24 kW prior to loading fuel.

temperatures, a maximum fuel clad unaffeded by fusi conditions inside the

29. Cornment. One commenter noted tempereture of 708 'F was calculated MSB. Helium was chosen because it is that Page 4-1 of the SER for the VSC-assuming an ambient temperature of inert and it has good best transfer 24 cask states that the applicant will 100 T. His temperature is 4 Y below characteristics. N fact that the F1. St.

remove any cask imm service which has an acceptable fuel clad tempereture Vrain reector used helium as a coolant inlets and outlets blocked. It should say cnterion of 712 T. He NRC ted did not r:ontribute to its operational "or" insteed of"and." this margin on the basis of the fol 'ng problems. Response. The statement refers to a conservative factors applied in the off.

28. Comment. One mmmenter wanted proposal made by the applicant and is normal case analyzed in the SAR:

clariScation of"approximately 24 kW." correct es quoted on peps 4-1. However, a ne calculation assumes steedy when referring to the hat somrta loaded the NRC did not accept this proposal state conditions. It would take several into the first MSB for tests mnducted by because the a licent did not provide days of sustained 100 T ambient the licensee to verify best removal a able ence that the cask will temperature to approech the calculated uPacity of the VSC systas. N be equately cooled in the event of a fuel clad temperature value of 708 T. commenter also indicated that the full blockage of either allinlets or

b. The fuel temperature criterion is Cert 1Scste of Compliance is weerly outleta. Sedians 1.3.1 and 1.3.4 of the based on prevention of fuel failures due restrictive in requirl a 24 kW best Certi5ceae of Comphance require that a to long-term degredation mechanisms.

load for the first cask some VCE be removed from service whenever Short term varistions in the everage reactors do not have fuel either aD inlets or all outlets are found temperature, such as when the daily assemblies which snake up the 24 to have blocksgo for 24 hours and the summer averege temperetures exceed 85 kW heet load.De ocnnmenter concrees temperature citerion of 350 T T. have no effect on tM long term reenmmanded that the requireement be has been exceeded. His ccmclusion is h ed to aim that the first cask be also stated on 4-1 of the SER. degradation mechanisms thet affect the 5 fuel cladding. Therefore, the annual loaded with a load as high as 30, rw' commenter noted sverage 75 T temperature wou}d be a Prer+imMa (but not to exceed 24 kW) to that Table 4.1-1 of the November 1991 more realistic condition to use in the veri the calculated heet removal SAR lar the VSC-24 cask falls to state ~ calculation than the 100 7 temperatur, cs liity. Another commenter asked what the tempemture difference would actually used in the calculation. why not test the czak with artiSenal be if allinlets were blocked over a long-

c. Heat conduction in the axial thermalloeds rati:er than with spent term.

direction is treated manerystively fuel. Response. & casamester is mrrect. because httle credit is taken for best Response.De intent of the language, However, a temperature criterion of transfer out of the ends of the MSB

• approximately 24 kW" was to provide 350 T bas been estabhabed for the canister.

some flexibility to a potential user concrete caek. Calculations indicate that

d. Fuel clad temperature is treeted because there is imo way to ensure that a temperature of 350 T could be conservatively because a peak beet the first fuel placed in the cask will reached aber 30 boors if either all inlets generstion rate rather than en everage have a heet loed of exactly 24 kW that or all outlets are blocked. lf this was used in the calculation.

wa.a used in the thermal analysia. The situation is identined. the licennae must Nse conservative lacsors uand in the purpose of the test is to measure the demonstrate that accident tempervt ure calculation of fuel clad tempero*ures cask performance and establish baseline criteria have not been exceeded or is provide reasonable assurance that the data. Folicnring loading and required to take the cask out of servv*

Federal Register / V:1. 58. No. 65 / Wednesday, April 7,1993 / Rules and Regulations 17957 l NRC notes that ruching 350 *F is not plan wp squimd.%ey also stated that and security requirements for an ISFS! an unsafe condition with respect to the there la a lack of contingency phnninir located within the owner contmiled containment integrity of the MSB or the for catastrophic events. ney noted aree nf a licensed reactor site. stored fuel. Rather it is a critwion for these events could include but would Section 72.212(b ros that the deciding whether to take the VCC out of not be limited to: spent fuelin the ISFSI protected l i service.His action is highly

a. Direct or indirect lightning strikes against the design basis threat for conservative, sina only the onset of on the casks.

radiological sabotage using provisions i very slow degradation occurs if the

b. Plane crash into the casks; and requimments comparable to those concrete temperature reaches 350 'F. As

c. Sabotage; applicable for other spent fuel at the discusad below,in twponw to

d. Earthquakes; associated reactor subject to certain Comment Number 31. a conservative

e. Fire; and additional conditions and exceptions adiabat.c heatup analysis determined

f. Emergency planning for cask desaibed in 10 CFR 72.212. Each utility that it w ould taka 7 days to reach malfunctions, licensed to have an ISFSI at its reactor unaccer table fuel clad temperatursa..

site is to develop security plans The NR : considers that within this time A commenter wanted the utility to notify either state or local ovemment and in a secunty system that before loading casks to make sure localProvides high assurance against frame.'.no licenses s enhanced daily survo!!ance prm; ram, which must servicae were aware and would know unauthorized activities which could include a comp =ent that veriBos the thermal perfore ace of the cask, would h w to rupond if muy undu tho constitute an unreasonable nsk to the public health and safety. The security identify the blot.iage and allow "p*, CI]}*"; Code of Fedwd g systema at an ISFSI and its associated suf5cient tune for necansary conective actions to be taken. Regulation.10 CFR parts 50 and 72 reactor are similar in design features to

31. Comment. One commenter requires that nuc! ear pknt structures

• ensum the detection and assessment of indicated that the safety evaluation for systems and components important to unauthorized activities. All alarm the tipover of the VCC only considered safety shall be des! ped and annunciations at the ISFSI are propriately protected against dynamic monitored by the escurity alarm stations apfects, including the effects of tornado-at the roector site. Res the structural aspects of the accideat e

and ignored the thermal consequent.es. h ed. Each ISFS pwlodiall The issue raised was that the VSC-14 driven missiles. that may result from inspected by NRC and annuall audifed cask uniquely requires a vertical events and conditions outside the by the licensee to ensure that the ""Cl**' wer unit.nis includes the orientation to adequately reruve heat effects o{possible airplane crashes. security systems are opwating within and that heat removalin the i Mzonts! configuration is degraded evr. if all ne licznsee s site evaluation for a their dosip limits. The validity of the threat is continually reviewed, with a nts blocked which au ld not at ta eons 1d* hrmkevaluation every six months by ,,,, d d Response. Dermal consequences of a activities. A conses proposing to use An adequate evacuation plan exists VSG-24 cask tipover were considered the VSC-24 cask is required to evaluate for the use of certi5ed casks because of and are bounded by the adiabatic heat, and verify that the SER for the facility the fact that the existing reactor up analysis performed for the cask. encompasses the desip basis analysis emergency plan covers the entire site. In Adiabatic heat-up is not affected by Performed for the VSG-24 or any addition, contingency planning for the orientation, eithw borizontal or vertical. certiSed cask. Genwally, a cask's events described above exists because ne adisbetic analysis determined that inherent design will withstand tomado these events are covered within the it would take approximately seven days miulles and other design loads and eme ncy plans of the reactor facilities to nach unacceptable fuel clad thus, also provides protection against whi will ues the cask. In accordance temperatures. The NRC considers that the collision forces imposed by light with to CFR 72.212(b), the reactor within this timeframe the licensee general aviation aircraft (i a.1500-2000 licensee must review the emerpncy would tale necessary corrective actions Pounds) which constitute the majority plan to ensure it provides adequate to return the cask to an upright position. of alreaft in operation tods. NUREG-protection. The licensee's emergency

32. Comment. One commenter stated 0800. Section 3.5.1.6 **Stan Review plan providae for responsive action if an that an analysis based on Diffusion Phm for 1.ight Water Reactors", contains event has happened which has the Controlled Cavity Growth (DCCC) he.s methods and acceptance critaria for possibility of creating an emergency or been the only method accepted by the determining if the probability of an after an actual emergency has occurred.

NRC to determine the maximum accident involving larger aircaft (both nrough communications between the allowable fuel cladding temperature. mittary and civilian) exceeds the utility and govemments, the contents of The commenter further stated that it

• CC8 Ptable criterion. It is incumbent the emerpacy plan and the actions to was not apparent that an analysis based upon the licensee to determine whether be executed by a ch entity for various on DCCG had been performed in or ne4 the reactor site parameters are situations are understood. In eddition, evaluating maximum cladding enveJoped by the cask desip basis as the utility is required to conduct a temperature for the VSC-24 cask.

required by 10 CFR 72.212[b)(3).nis periodic emergency exercise involving Response.The NRC 35rees that DCCG wouldinclude an evaluation the utility and govemment agency staff. is the only current method acceptable to demonstrating that the requirements of

34. Comment.One commentw stated the NRC to determine maximum

$ 72.106 have been met. that there was no contingency for allowable fuel clad temperature. The NRC reviewed potentialissues related accidents except to reload the spent fuel VSC-24 cask wu evaluated by this to Possible radiological sabotage of back into the cooling pool which may method. See Section 5.3.3 of the SER. storage casks at reactor site independent not he possible due to lack of pool E. A number of commenters expressed spent fuel storage installations (ISFSis) storage specs orignpact on the spent concern about emergencyplanning and in the 1990 rulemaking that added feel due to the acx:ident. response to contingencies, subparts K and L to 10 CFR part 72 (55 Response.Because of the design

33. Comment. Some commenters FR 2918 t). NRC regulations in 10 CFR features, as well as the procedures and expressed concern that no evacuation part 72 establish physical protection requirements discussed elsewhere in

17958 Federal Register / Vol. 58. No. 65 / Wm4===4=y. April 7.1993 / Rules and Regulations this response and the -r4ded safety cxrmmenters stated that the VL24 had systems curroody in operation.and analyris, the likelihood of an acx:Ident not been tested to the full range of during the se L.uon of the first See occurring which will require removal of climatic conditions. cusks, thet are expeded to k pleoed in the spent fuel from the cask is very Response. Although the volume of servics. Another question was raised small. However, even if such an data that is available to suppwt pointing out that tne vendor did not use unlikely accident occurs, the cask certiScation of the VSC-24 cask does weld inspectors lined /cortiBed to design is required to have capability to not include results of full scale tests, the American Weld ety D.1.1. permit retrieval. (10 CFR 72.122(1)). svailable data is more than sufficient to Response.b NRC ensures NRC does not require a licensee to show that the use of the VSC-24 cask by complianos with 10 CFR 72.236 (D and maintain a reserve capability in the a licensee will not place power plant (k) through inspections, and ensures spent fuel pool. Many licensees may do workers, the public, or the scrironment compliance with to CPR 72.236 (1) and so, however, and they would, therefore. at any undue risk. Also the mnditions (m)through the cask op royal prnraan have the option of mturning the fuel to of use for the VSC-24 cank in the his process willidenti di!!orent areas the poolin the unlikely event of an Certincate of Compliance ensure that may need correcti. but that is the accident requiring removal of fuel from adquate protection of the workers, the of an inspection program. lf a the cask. In addition, licensees will have puhuc. and the envirr=-nt. Further, vio tion of the requirementsis other c'ptions available to cover this the VSC-24 cask has been desipW ed and detected, b NRC can impose

alties, unlikely mntinBency including will be fabricated to well" oman stop work.b NRC

• s notei temporary storage in a spare storege criterie of the ASME BAPV and ACI of the fact that problems noted by b cask or use of an existing certined codes. In addition. it uses construcion commenters were identi5ed as a result transportation caak. Licensees would materials which have well known and of NRC's inspection progrom during b teve io consider these. and other documented propert>es to provide the construction of specLSc casks.His available options,in the unlikely event normary structural strength and experience reemphasizes the need for an accident occurs requiring removal of radiation shielding to meet regulatory close and continuing quality the fuel.

requirements. While the NRC nas not surveillance under vendor and user QA F. Other comments which do not relied on testing of the VL17 cask (a programs during all VSC-24 and other specificallyfit those categorww above smaller version of the VSC-24 cask cask mostruction activities.N NRC follow below. These comments deal with design) for approval of the VSC-24 cask, will continue to conduct the inspections a brood range of other technicaland b VSC-17 cask has ben tested by of construction activities in accordance procedurnt issues. DOE at its Idaho National Engineering with NRC's Inspection ihdut In

35. Comment. Nro are outstanding Laboratory. N report **Performanos conjunction with vendor's quahty safety issues that the NRC arpects to Tesdng and Analysis of the VSC-17 assurance (QAl program, specl5 cations.

resolve in the Erst test. Ventilated Concrete Cask." EPRI TR-drawings, etc. to ensure quality work. Response, b NRC SER addresses all 100305, dated May 1992, concluded that As to the specinc point of the signi5 cant safety issusa, and there are the VSC-17 cask can be safaly used at quallBestitm of welds and inspedors, no outstanding safety issues about the reactor sites. While the VSC-24 cask the NRC notes that the welds referenced VSC-24 cask that remain unresolved. approval does not raly on the VSC-17 were not structural welds and, as Acmrdmgly, the first test does not cask, the desips are almilar and many allowed by the vendor's fabrication involve any safety issue. !ts purpose, parallels in design and function can be spedEations, do not have to be rather, is to benchmark the best removal drawn. DOE testing of the VSC-17 qual 15ed to the same extent as a capability of the VSC-24 cask. demonstrates that ventilated storage structural weld.

35. Comment. One commentar asked cask technology can provide safe storage

40. Comment. Concern vu expressed that a requirernent to subtnit a report to of spent fuel nus,in view of the that the measuremmt of actual i

the NRC within 15 days of the test and above, although the commenter's effectiveness of a Mi+;y in evaluation of the Erst cask and prior to observation that the VSC-24 had not - delivering stated requirements must be construction of the somnd cask be been fully tested under climatic d=aam ted empirically, and that the i s added to h VSC-24 cask Cartdicate of conditions is technically correct. the NRC has not demonstrated the goal of Cornpliance. Also the report and cask has been designed for ambient this technology defined acceptance sub ant NRC veview should be tmym.w. extremes from -40 7 to criteria. or speci5ed how complianos is p! in NRC's Public Document +100 7 and masts the ASME and AC2 demonstrated. Some commenters also Room. requirements. expressed concara that the review of the Response. A letter report summarizing

38. Cornmerrt. One commenter noted concrete cask was not done et the sarne the results of the thermal test and that Consumers Power does not have a level as that performed for metal casks evaluation of the first cask placed in plan to remove spent fuel stored under and that no independent computer i

servicz will be submitted to b NRC general license from the reector site as analyses were for the design j and placed in the Pubbe Document required by to CFR 72.218. event review. commenters noted Room.m licensee may, at their own sesponse. W licanese is not required that the rrriew requires more than j financial riak, fabricate additional czaks to have a plan to remove spent fuel limited courpuner models. prior to using the first cask. If h first stored on site under the general liosese Respa=.For the leeue of acceptance cask does not perform as specified, the until en application to terminate the criteria, the NRC has establiebed NRC would prevent use of the other reactor operating license is eubmitted to speci8c requirmants in 10 CFR part 72 casts or modify cxeditions on how they the NRC. This requiremeest is found in that asuet be met in order to obtain a could be used. to CFR 72.218(b) and to CFR 50.54(bb). CertiSoste of Cosephena for a cask. W

37. Comment. It is unacceptable from

39. CommerW.One commseter noted details of the reeww and bases for the a public health sad asisty standpoint to that the NRC does not speci5cally NRC conduding that b cask meets b j

conduct the first full sesle test of a VSC-require inspections against to CPR requirements of to CFR part 72 are 24 cask at a reactor site because it pieces 72.236(j)-(m). Questions won raised provided in b SER.b 0*I of df7 E the power plant workers, the pubhc. regarding quality assurancs problems cask storege 6.hseleir is to store spent and the environment at risk. Two encountered during the inspection of fuel safely.Nt goal, and the

Federal F"6 a'r / Vol. 58. No. 65 / Wedneedsy. April yo 1993 / Rules and Regn1=H-17959 h effectivenees of the technology, concrete modale desips la the Snel have connA-= in the analyses which prvvi has been demonstrated subpart K rule: are done or if b design relies on empirl y and experimemtaUy. A mm)ar reason that these snent foal storage nationally recognised codes and Different cask designs may require systems le4. NLSKMS, heoduler Vault Dry standards. Testing to destruction is an dif!arent types of analysis to storel. winch an being considered by the option that can be used to conErm demonstrate their safety. an.d therefore Commission br use under e pseeralI-design adaquacy. However, destructivo different review methods may be are not beirg approved at this time is that appropnate to reach that conclusion. In they have compents that are dependent on tests of an entire cask are not necessary each case the level of review performed swspsanc paremers and, thus, to evaluate a design when other non-is that needed to provide amurance of $8PacSC 8PProvals. 55 FR 29181 ulyt8. destructive tests or destructive test g the componwats wiU provide the adequate protection of the public health necenary information to evaluate e and safety. M reover, the NPRM included the

41. Commer f. Some commenters staternent that "(t)he Commission has desip.

43. Coenment Some commenters claimed that part 72. subpart K was evaluated and approved. in spectSc

,rpnosed ca==us that fuel handling orip'r. ally intended to apply to metal licenses issued under 10 CFR part 72 could be underless than ideal casas only. Concrete task systems were other types of dry storage modules (and conditions and bt storap could be not addressed in the original t)hese mebds may be appmved in the under harsh environmental conditions. rulemaking. future for use under a general license." Sites where the VSC-24 cask is Response. As discussed below both 54 FR 19362. It also noted that pmposed far use would experience low the languay and history of subpart K "(s)torage casks certi5ed in the futum winter temperatures, freeze-thew cycles, show that it applies to any NRC. will be routinely added to the listing in high humidity, and marine conditions. approved dry cask storap system $ 72.214 through rulemaking Concern was also erpressed that harsh including concrete casks systems, and procedures." 54 FR 19380. environmental cond tions and damep commenters are therefore mistaken in These statements collectively show tc the MSB protective coating will their view that it was intended for metal the Commission specifically envisioned degrede the containers as a result of casks only. the possibility of future rulemaking (i.e.. corrosion embrittlement, cracks. fatigue Subpart K applies "to casks appmved the proceduto NRCis now using) to add and other aging effects which would under the provisions of this part" which concrete storap systems to the list of afled the ability of the cask to survive includes casks approved by NRC under epproved spent fuel storap casks in e,,r extended periods. to CMt part 72 subpart L. Subpart L subpart K. Consequently. concrete Response. Handling of fuel and contains NRC's approval conditions "for storege systems can be " casks appmved loading of the cask is performed under NRC spent fuel storage casks designs" under the pmvisions of this put" for well controlled conditions in 6 which would include concrete casks. purposes of part 72. subpart K if, for roector's fuel handling building using None of the approval conditions in example, they are not dependent on written procedures developed in subpart L requins that the cask must site-specinc parameters and therefore do accordance with the reactor operating use a metal cask design. not require site-specific appmvals and if licsnse. De VSC-24 system has been Additionally, there is information on they conform to the approval conditions evaluated for the possible effects cf concnte stornp technologies in the of subpart L. harsh erwironmental conditions and the subpart K rulemaking record that would Finally,it is noteworthy that the MSB has been evaluated for the possible caska.pport limiting it only to metal Commission adopted subparts X and L effects of cormsion due to humid and not su Spedfically, the Commission's for the express purpose of implementing marine environmental con &tions. As a notice of proposed rulemaking (NPRM) certain interim storego provisions of the result of the corrosion analysis of the for subpart K nferenced the Canadiar.s* Nuclear Waste Policy Ad of1982 that. MSB.the NRC found the design use of" concrete casks called silos"in significantly, es not limited to metal acceptable with b consideration of dosenbing "the knowledp and cad a 54 FR 19379 (May 5.1989). In localized corrosion mechanisms (i c., experience of dry spent fuel storage in particular,the Ad authorised the pitting, stress cormsion cracking. concrete casts." 54 FR 12379-80 (May Commissloo to opprove by rule "one or cnvice mrrosion and galvanic

5. t989).ne proposed rule also mces (storage) technologies" for use et mrrosion) as well as general corrosion.

referecced DOE's demonstration of dry roedor sites. (Sec. 218(a) (42 U.S.C. Localized cormsive attack on the MSB storage in saaled storage casks (SSC) 10198(a)).%e Act also directed the surfaces is minimized by choice of which it described as "an abcve ground. Commission to establish procedures for meterials and design fostures sur.h as steel-lined, reinforced concrete cylinder the licensing of"any technology" the ceramic tiles between the VCC hner or catk." Id. Further. it cited experients spproved by the Commission under and the bottom surfoce cf the MSB. gaiced fro:n spent fuel storage "in section 218. (Sec.133 (42 U.S.C. Furthermore, the NRC allows no cred t stainless stent canisters stored inside $ 10153)). nerefore, beasuse the Act's for the attributes of the paint. concrets modules at the H.B. Robinson provisions are not limited only to metal Aging issues attributed to fatigue for 2 site * * * "Id. lf the Commission storep caak designs. it would be the MSB woro evaluated accordirg to had intended to limit subpart K to metal inconsistent with the Commission's the ASME B&FV Code. Section !!I. e nd casks. It would not have included data purpose to limit the application of it met acceptable standards. fmm other dry storage technologies in subparts K and L to such designs. Tempersture extremes, such as free:e. the record supportirig its action.

42. Comment. One corasnooter thew cycles which exist in the Grnt Although the Commission has no' requested b proceeding be stopped Lakes region, were considered in the 1

previously approved mncrete storage until the NRC rerises all regulatory evaluation of the VSC-24 cask. j systems (or casks) under subpart L. it requirements pertaining to the storego of According to the conditions fer ead j expressly noted sucf systems might be high. level weets and spent fuel to use.the user of the VSC-24 system mli j approved (and thereby included in require testing procedures which Perform site-specific analyses to u

• v subpart K)in the future. In particular.

include testing to destruction. that the temperature conditions 1 the Commission gave the following Berponse. no NRC dose not require assumed in the analysis bound the explanatim for not appmving certain testing to destruction or other tests if we conditions existing at the site l

17960 Federal Register / Vol. 58, No. 65 / Wednesday, April 7,1993 / Rules and Regulations N possibility of MTC and MSB the radioactive fuelloaded into the failures"and the language which the cracks was addressed as a part of ferritic MSB. However, an additional margin of NRC uses in Table 1-1. Characteristics material considerations. Based on safety is provided because:(1) ne of Spent Fuel to be Stored in the VSC-guidance provided in ANSI N14.6 and welded joint is a double weld;(2) the 24 System, referring to Fuel Cladding as: NURSG CR-1915 tha NRC established weld joint hu been analysed acx:ording "Zircaloy clad fuel with no known or test and operating limits for the MTC to ASME Section III criteria for all load suspected gross cladding failures." and the MSB to preclude the possibility conditions including accidental drop; Response.no NRC agmes that there of bnttle fiacture. (3) the pressure inside the canister is an inconsistency. Acuptability is Ennfly, tha VCC is designed and during normal storage operations is based on zircaloy clad fuel with no f.tr cated to Amarican Concrete epproximately atmospheric, resulting in known or suspected gross cladding bnute Cok requirements which very low stress intensities; and (4) the failures. Soction 1.2.1 of the Certificate ar. suer durability under extreme confinement integrity is established by of Compliance has been revised to con 6tions for extended periods. The ASME code test procedurse, which

• • specify no known or suspected gross cask is also sub}ect to annual visual include dye penetrant testing of the root cladding failures."h intent of this surface inspections for chipping, and cover welds of both the inner and specification is to rely on the cladding spalling, or other surface defects. Any outer welds. In addition, the NRCis to safely confine the UOs fuel material surface defects found can be easily requiring testing for helium leaks prior within the rods to preclude operational corrected. The fluence of the neutron to the placing of the MSB la storage.

safety problems during its removal from flux within the spent fuelis five orders

46. Comment. A number of storage. Fuel cladding with pin hole of magnitude less than the fluence commenters questioned the lack of leaks is still capable of confining the encountered within an operating transportability of casks and the fuel and therefom is acceptable for reactor, and therefore embrittlement of apparent noncompliance with the storage in addition the inert atmosphere the MSB is not of concern.

requirement of 10 CFR 72.236(m). and fuel clad initial temperatures

44. Comment. A commenter asked Several commenters expressed concern provide assurance that the cladding will how the NRC will correct the problem that the VSC-24 cask is not compatible be protected during storage against when something goes wrong with the with transportation requirements.

degradation that leads to gross rupture. VSC-24 cask. In the event of a tipover Several commenters questioned how the

48. Comment. Commenters stated that or drop of a loaded VCC. the comrnenter spent fuel will be transponed to a there is no evidence that PSN believes the licensee should be required Federal Repository and what will be the considered the effects of worst use to report the incident to the NRC within additional handling cost.

tolerana combinations in the structural 4 hours and the NRC. rather than the Response. These casks are cvtrently analyals. licensee. should determine whether the approved for storage of spent fuel. not Response.nere are several generic MSB and/or the VCC should be reloaded off-site transportation. Therefore, there areas where improper tolerance for spent fuel storage. is no need for the VSC-24 cask to be combinations could leopardiz.e the Respense. The hcansee is responsible compatible with transportation structuralintegrity of a design. These for correcting problems when they requirements.These casks are only areas are: occur.The NRC is responsible for moved between the fuel handling (1) Over-tolerance of weight which ensuring that the licenses takes building and the storage pad at the site could result in unallowable strus levels appropriate corrective action. Dese where the fuel will be stored. Although for some components; rules reflect existing regulatory practice to CFR 72.236(m) states "To the extent (2) Improper tolerances for dynamic and procedure. The regulations and practicable in the design of storage parts such as in machinery which could Certificate of Compliance identify casks, consideration should be given to result in interference and failure; specific events and conditions where compatibility with removal of the stored (3) Improper tolerance for fuel the licensee would have to notify the spent fuel from a reactor site, positioning in the basket: NRC. transportation, and ultimate disposition (4)Irnproper tolerances of parts of an In accordance with to CFR 72.216(a) by the Department of Energy." there is assembly wnich could lead to induced the licensee is required to report cases no requirement that the sto.ege cask stressee frora an interference fit or the involving any defect as a result of a itself be transportable off site. If the cask converse situation, i.e., loose tolerances tipover or a drop to the NRC within 4 vendor wants to have its cask used for which could lead to an ill-defined load hours. The licensee would also have to the transportation of spent fuel,it would path; and inspect and evaluate the MSB after an have to obtain a transportation (5) Improper tolerances which might tipover or drop of 18 inches or higher.y Certificate of Compliance issued by the cause a heat conduction path to exist or Based on that evaluation, the licensee, NRC under to CFRpart 71. not exist. not the NRC, would be responsible for %e mechardsm for transporting the Re NRC has reviewed and venfied determining continued tise of that cask. spent fuel from a reactor site to a that tolerancas spec 16ed in the NRC's responsibility is to monitor and Federal Repository is unknown at this application would prohibit a wemht oversee the licensee's activities. NRC time. However, it could be by truck, rail. which is above the loed used in the has, however, the authority to order the barge, or some combination. Also, the calculation package,no NRC also licenses to cease use of a cask,if that handling costs an unknown since DOE reeiewed speci6ed dimensioning. were determined to be necessary. compatibility requirements are not which, when followed as required. e,l

45. Comment. One commenter stated known and regulatory requirements at prohibit interference and failure of that the double seal welds at the top of the time of transfer could be different.

dynamic parts such as machinery or fel the MSB do not comp ly with the ASME

47. Comment. One commentee positioning in the basket.no NRC Code, Section III, Subsection NC.

pointed out the NRC indicates that the nyiewed the vendor's calculations to Response.De double seal welds at analyses presented in the SAR are assure that the loads which were the top of the MSB meet all of the ASME " based on non consolidated, zirtaloy. analyzed and best conduction paths requirements except the volumetric clad fuel with no cladding failures." account for the range of tolerances F r inspection requirement. 'ais inspection Please clarify whether then exists an these reasons, the NRC has concie i ) is not possible due to the presence of inconsistency between "no cladding that tolerance combinations are l i

6 Federal rapaar / Vol. 58. No. 65 / Warin==itay. April 7,1993 / Rules and Regulations 17961 adequately addreemd for the vendor's to produce enough bc Itng for optimum wt.% Usss ir. ' w spent bl pool were strudural and thermal analysis. moderation. brefors.NRC would pleoed in the mSB.

49. Comment. A comme:rter indicated accept a L of 0.98 for any mek Response. Tes NRC agrees that the that the VSC-24 was exempted from generically for this accident case, but a baron speciScation in b Cwtiacate of established dadding temperature L af a 95 would apply otherwise.&

Compliance for the VSG-24 cuk may be criteria for short arm normal condition conditions of nudear critimhty, and the restrictive. h bcron speciacation is events. in which the maximum fuel experiments that provide that consistent with the roaximum allowable cladding temperature limit is exceeded information can be and have been uranium enrichment (4.2 wt.%). based by as much as 170 'F. measund with a high degree of on the criticality analysis presented in Response. The VSC 14 has not been accuracy since the 1940's. N age of the SAR.b Certificate of Compliance exempted from a short term temperatur* the data is not a.g:uScant. !! is desirable specittcation for boren concentration in limit for fuel daddmg. In comparing the short-term and long-term thermal that the benchmark experiments water is a bounding condition which represent h system under evaluation was chosen to limit reliance on hydraulic evaluation shown in Tabl* as closely as possible.N features or administrative controls to determine the 4.1-1 of the SAR. the short-term parameters that are irnpartant to this Pro requimd boron concentration for temperature will exceed the long-term. purpose are the bl compositten and cask loeding. A method like that temparature by as much as 170 'F. This higher temperature howevw is enrichment.the pometry of the fuel Propmd by the commenter, to assambly. i.e., rod diameter and pitch, determine the boron concentration acceptable during the short term while the..aalis dried prior to Elling h MSB dadding type, and any neutron required, based on the maximum initial with an inert gas (helium). weld wohng absorbers in the vicinity of the fuel pina. Um enrichment of fuel at each nador the MSB and final plamment of the These parametars must be properly sits could be considered as a future MSB in the cast for mtwim storage.N considered in the pre==aing of nuclear amendment to the Certificate of NRC conserystively assumed that air cross sections used in criticality Compliance. analysis so that the benchmark

52. Comment. Some commenters was present during the drain-down and dry-out periods and calculated the experiments an used to determine e sugpstod that the NRC should consider oxidation rata. h maximum length of method bias, or systematic error that limiting the cask storsp time and fuel oxidation for defective fuel was mey result from the particular set of

• rPnned concem that cask storage determined. h cladding strain was nuclear cross section data that are used, could bomme permanent if the DOE might not acce required to do.pt fuel as they are or from b methods used to rocess the estimated to be lees than 1 percent.

Commenters also noted con actbn dau. Onm d bin is Therefore no defect extension or fuel powdering is anticipated. N abort determined for the Particular fuel dat the NRC requirement that cask viabill be evaluated for "at least" 20 terra mcmased temperature is desirable peamm b ulcu g i to ensure removal of moisture. inedve m & mic pury years, es not,in itself. guarantee safety in the appenntly likely event the g "Y,am:t is not necessary that the Following dry-out and helium nefm. I. caaks remain years or decades beyond introduction, the temperature will' drop the originalintended duration. below the long term limiL gmsa w macmacepic geornetry of the Response. By approval of the

50. Comment. Some of the bechmark experiments be similar to CertiScate of Compliance, the NRC has commentes indicated that the SER for the VSC design as long as the method limited the cask storege time to 20 years.

the VSC-24 caak allows L of 0.98 and bias has been determined for the After the 20-year period, the certificate that this deviates from the normally 8PPropriate fuel parameters. The B&W can be renewed, with each renewal accepted limit of 0.95 speciSed in NRC citical experiments have been widely period not to exceed 20 years, upon Regulatory Guide 1.13. Proposed used for this purpou since they war

• demonstratim of continued protection Revision 2." Spent Fuel Storage Facility Performed using light water resc2cr fuel of the public health and safety and b Design Basis." ne commenter assemblies similar to those used in environment. In the event that safe indicated that NRC abould allow other many light water reactws-storage of fuelin a particular cask vendors to modify their cask to 6 of

51. Comment. One commenter r=nnatbe anstrated beyond 20 0.98. One commenter expressed concurrn indicated that the CertiScalm of an alternate means of storage will that the beochmark experiments that Compliance for the VSC-24 osak is required. Finally, DOE is required by won cited in the analysis dated to the unnecessarily restrictive in requirint b Nuclear Weste Policy Act of 1982 to 1970's and because of their age were that the MSB contain 2A50 ppm boron aampt spent fuel for Wtimate disposal.

considered insppropriate for use and solution while it is being loaded. This As one cx>mmanter notr d. DOE is commented that there was a difference concentration of boron would keep 6 pmposing a new strategy in which in the geometry between the benchmark less than 0.95 even if all 24 storege Congress would authertre it to select a calculations and the VSC-24. specas in 6 MSB were loaded with site in time to recalve spent fuel for Response. The 6 of 0.95 is guidance fuel assernblies which average 4.2 interim storop 1998. and is thus, not a requirement. As such, weight percent (wt.%) Uzis. Some

53. Comment menters indicated a licenses has flexibility and may nuclear power plants do not have 4.2 that PSN made as error in calculating propose an alternative limit. Based upon wt.% Um fuel on site. Herefore, there the does rete et the gap between the NRC review. NRC acx:epted the is no posaihility of fuel mataining that MSB and MTCL PSN had 440 mrem /hr licaneee's proposed up of a L of 0.98 concentration of Um being loaded in a compared to NRC's calculated 4140 for the aa;ident came of misloading the MSB. De crimmseter recx>mmanded mresa/br. Why weren't these MSB with all fresh fuel of marimum that the Certi$cate of CornpH-discopancies resolved? How would ennchment and optimum modarotion requirement for boron cane==tration in welders be protectedf conditions. His acx:ident condition the MSB cavity water be changed to Aespoems.PSN did not make an ermr bordars on the incredible sinc = 11 allow other concentr*Hans to be used in their at~1 Haa Rather, they made requires a mutually exclualvo condition: such that the boma concentration used an error when transcribing e calculated that is. 24 unfiradiated fual anemblies would maintain L less than 0.95 even value la en SAR table. This discrepancy ht have beat generation rates suf5cient if fuel anamblies containing the highest is identified and neolved in the SER (pg

i 1 17962 rederal Register / Vol. Sa, No. 65 / Wedneedsy, April 7,1993 / Rules and Regulations { 6-12). With twped to protection of

56. Comment. Commenters believed to elicit and fully canalder public welders. 6 operating procedures and that PSN W=-

of shielding comments on the VSC-24 technology, radiation protection program of the codes against m dose reise for Section 133 of the Nuclear Waste licensee willinclude precautions so that the VSC-24 cask was grossly in error. Policy Act of1982 authorises NRC to the exposun of personnel working with Further. PSN did not benchmark b approve spent fuel storage technologies the system inside the fuel handling SKYSHINE-U calculation method.The by rulemaking. When it adopted h building will be maintainedwithin 6 NRC calculated direct and air-scattered generic process in 1990 for review and dose limits of to CPR part 20. dose rates, at various distances im;n b approval of dry cask storage

54. Comment. Commenters stated that cask, which were many times higher technologies, the Commission stated the nported dose of 130 mrem /hr for than the PSN calculated does rates.

that " casas * * * (are to) be approved the VSC-24 cask sides is still 6 times Response. PSN's benchmarking of the byrulemakm and any safety lasues that higher than the stated limit / ANISN and QAD computer codes for are conn with the casks are specification of 20 miem/hr. dose rate calculations was found by the Properly addroened in tlat rulemaking Response.He limit of 20 mrem /hr NRC to be incomplete because it did not rather than in a hearing procmdure." 55 stated in section 1.2.4 of the Certificate address differences in does rates FR 29181 Quly 18.1990). Rulemaking of Compliance applies to the sides of the calculated by the ANISN and QAD under NRC rules of rectice, described VCC, at the pad. The 130 mrem /hr value computer codes. no NRC conducted in to CFR 2.804 an 2.805, pmvides full quoted in the comment refers to the independent confirmatory calculations opportunity for expression of public maximum dose rate et the sides of the to estimate the dose levels associated views,but does not use formal hearings MTC when loaded with the MSB,inside with the VSC-24 cask system for of the, type requested by commenters. the fuel handling building. Because the comparison with the vendor's In uns proceeding. rulemaking clearly MSB has not been loaded into the VCC calculations. Based on that comparison, Pmvided adequate evenues for members cask at this point. It is not subject to the the NRC concluded the design provided of the public to provide their views 20 mrem /hr speci6 cation. ecceptable shielding, regardmg NRC's proposed spyroval of

55. Comment. Commenters believed Although PSN did not benchmark the the VSC-24 cask, including we that PSN made several mistakes in SKYSHINFAI calculation method, they Opportunity to participate through the submission of statements,information, how much radiation might used that method to calculate site calculatinfe surface of the VSC-24 cask.

come off t boundary dose rates. Based on review of data, opinions and arguments. In this Because the VSC-24 cask has never their calculations and inde dont NRC c nnectim,the NRC staff prepared for tion Pa been built,it is fair to say that no one calculations, the NRC conc ded that Q ,t ms fw 6. 24 has any definite idea of what the actual PSN had not calculated conorrvative dry cask stem each time making dose rates will be. In addition, some neutron and gamma dose rates at the detailed.kocum,ented findings of commenters noted that conclusions site boundary. However, even with the compliancs with NRC sefety, secunty drawn from testing a pmtotype are of. NRC's more conservat!vely calculated and environmental requirements. The dubious import "when dealing with the site boundary dose rates, the NRC staff's first evaluation. d in effects of radiation." concluded that general licensees using March 1991, reviewed an proved the Response. As stated in section 6.3 of the VSC-24 cask will meet all VSC-24 for referena in a site-specific the SER, e number of errors were applicable regulatory requirements. application for an independent spent discovered in the vendor's shielding in addition, the NRC also requires any fuel storage installation. in May 1992, analysis. An adequate explanation for VSC-24 user to measure the external the NRC staff reviewed the VSC-24. and those errors was offered by the vender. cask surface dose rates to ensure the approved the design for purposes of However, the NRC made independent cask has been properly loaded and initiating this rulemaking to grant a confirmatory calculations to estimste rediation monitoring to ensure generic approval of the desi n. In 6 the dose levels associated with the complianca with regulatory addition, the staff conducted a third VSC-24 system.ne vendor's shielding requirements. review in response to the public design and expected dose rates along

57. Comment. A number of mmments on the VSC-24 in this the surface of the VCC were determined commenters requested a public hearing rulemaking, again Ending compliance to be acceptable based on a comparison on this rulemaking. Appmximately half with NRC requirements as set forth in with the independent NRC calculations, of the commenters requested that a full this notics of final rule and response to NRC agrees with the commenter that the public hearing be held at each reactor commenta.

actual dose rates from specific fuel facility site prior to the use of dry cask In addition to reviewing loaded into the cask cannot be exactly storsgo at that site. systematically and in depth the determined a priori. However, dose Response. Conaistent with the technical lasues important to protecting calculations can readily predict applicable procedure, the NRC does not public health, and the expected dose rates for the VSC-24 caak intend to hold formal pubb hearings on environment, the has taken extni with sufficient accuracy to assure that the VSC-24 cask rule or separste steps to obtain and fully consider public NRC limits will not be exceeded. In hearings at each reactor site prior to use views on the VSC-24 technology, and adclition, these calculations tend to be of the dry cask technology approved by has made every effort to rwpond to conserystive and tend to overwtimate the Commission in this rela-Har public mnewns ano questions about the actual done rates that would be Rulemaking procedures, used by the VSC-24 czak's compliance with NRC and environmental experienced during actual operations. NRC for generic approval of the VSC-24 safety, security %e initial public Prototype testing was not used in czak, including the underlying staff requirements. evaluation of the adequ of the shield technical nrviews and the opportunity comment period opened on June 26. design for the VSC-24 Finally, the for public input, are more than adequate 1992, and closed on September 9,1992. licensee will conduct surveys to ensure to obtain public input and assure In addition, NRC rocsived a number of compliance with regulatory protection of the public heahh, safety comments after the close of that period. requirements and the Certi5cate of and the environment. Further,in this all of which were fully considered. Compliance. rulemaking. NRC has taken extra steps Subesquently. NRC extended the period

Federal Register / Vcl. 58. N3. 65 / Wednesday April 7.1993 / Rules and Regulations 17963 for submission of public comments until Response.no NRC granted Pacifle stormse at a ranMar site is also b February 22,1991. nua. b public Sierra Nuclear Associates' request for an croclusion of other NRCEA's for comment pniod for this rule has exemption to fabricate a limited number previously approved dry casks analyzed effectively been almost nine months. In of the casks before issuance of the in earlier ruta==W= addrusing part addition. the NRC staff made every CertiScate of Compliance under its NRC 72 and in b Commission's Waste effort to consider comments received appmved quality assurancs program. Con $dence decisions in 1964 (August aner February 22.1993. Further, the and at its financial risk.De NRC's 31,1964: 49 FR 34658) and 1989 staff pmposed and participated in a finding. based on b SAR for b VSC- (September 29,1989: 54 FR 39765). In pubhc meeting near one of the nuclear 24 cask and 6 NRC's SER, concluded the 1964 Waste ConBdencs decision, the plants proposing to use the VSG-24 cask that beginning fabrication prior to the Commisalce concluded there was (i.e.. Palisades). with the Attorney issuance of 6 Certificate of reasonable assuranos apent fuel can be General of the State of Michigan, to Compliance would poes no undue risk safely stored at reactor sites without provide further oppmtunity for public to pubhc health and safety. Un of these significant environmentalimpacts for at input on the safety, security and casks is dependent on satisfactory least 30 years beyond expiration of NRC environmental compliancs issues in this completion of NRC's certi5 cation reactor operating licenses.b 1989 rulemaking. NRC also participated in an process. Waste ConBdencs decision review earlier meeting of the Van Buren County

61. Comment. Some commenters reafHrmed prior Commission Commluion near the plant site requested that the NRC prepare an conclusions on the absence of Under these circumstances. formal environmental imped statement (EIS) signi!Icant environmental impacts.

hear!ngs would not appreciably add to and update the Generic DS for the Thus, given the Commission's specific NRC's efforts to ensure adequate handhng and storego of spect fuel.N consideretion of b environmental protection of public health, safety and DS should be submitted to the impacts of dry storege summarized the environment, and are unnecessary to Environmental Protection Agency (EPA) above, and iven the sheence of any new 8 NRC's full understanding and and to the State of Michigan. Some information casting doubt on the consideration of public views on the commenters also requested that action conclusion that such impacts are VSC-24 cask. on this rule be delayed until the expected to be extremely small and not

58. Comment. Commenters believed Wisconsin EnvironmentalImpact environmentally significant, no that a full demoostic process is needed Statement is complete.

meaningful environmental insights are in this decision. Response.N potential. likely to be gained from further 1 Rerponse. Because this rulemaking environmental impacts of utilities using p tion of either en US or an was conducted pursuant to b the VSC-24 cask (or any of b other ted GEIS for the dry storage u procedures for approving dry storage spent fuel casks sporoved by NRC (10 methodology. casks for use under a generallicenw. as CFR 72.214)) have been fully considered he EA covering the proposed rule, as required by Congress in the Nuclear and are documented in e published well as the finding of no significant Waste Policy Act of 1982, and pursuant Environmental Aamment (EA) imped (FONSI) and published to 6 public notice and comment covertog this rulemaking. Further, as for this ruld 5. fully comply with I procedurw of the Administrative described below, the EA indicates that the NRC environmental regulations in Procedurw Act, the resulting final rule uw of che casks would not have to CFR part St. Moreover, since the approving the VSC-24 cask is the significant environmentalimpacts. h=6 ton's environmental product of a process prescribed by law. Speci$cally, the EA notes the 30 plus regulations in part 51 implement NEPA

59. Comment. One commenter stated that the gap between the MSB and the years of experience with dry storap of and give proper consideration to 'be spent fuel,identi$es the previous guidelines of CEQ, they assure that the MTC is given as 0.5 inch in WEP-extensive NRC analyses and findings EA and the PONSI cx nform to NEPA 109.001.4 and as 1.0 inch in Figure 5-that the environmentalimpacts of dry procedural requirements, and that 5 of WEP-109.w13. His comtrenter storage are small, and succinctly further analysse are therefore not legally also stated that the dose rate was not deschs what impacts there are, reguired.

i clear. inclu - the non-radiologicalimpeds m connedian with & EA and Respor se.b difference in the ofcask cation (i.e., the impacts IDNSI. it beare emphaatring that 10 referenced gap size is a consequence of associated with the relatively small CFR part 72, subpart K already chanpa made as a result of earlier amounts of steel, ccocrote and plastic authorises dry cask storage and tireedy reviews.no final design was based on used in the casks are expected to be sPProves dry casks for une by utilities to the 0.5 inch gap as indicated in WEP-insigni5 cant). the radiological impacts store ePent fuel at reactor sites. See to 109.001.4. De calculation of WEP-of cask operations (i.e., the incomental CFR 72.214 for a listing ofinfortnation 009.0013, which uses a 1.0 inch gap,is offsits doses are expected to be a small on Cask Certificate Nos.1000 through { i therefore conservative for shielding fraction of and weil within the 25 2003, h present rulemaking is calculations. Because the gamma dose is mism/yr limits in NRC regulations), the eczordingly for the limited purpose of j more than 30 times that due to neutrons, potential impacts of a possible dry cask adding one more cask to the list of casks any small decrease in the neutron does accident (i.e., the impacts are expeded already approved by NRC. Furthermore, rate, due to a smaller gap, would not to be no greater than the impacts of an the ask, to be added to the NRC list by signiacantly change the calculated accident involving the spent fuel storap this rul will comply with all neutrou and gamma dose rates used to basin), and the potentiafimpacts due to ap licable safet assess ocrupational exposure. In possible sabotage (i.e., the offsite does is inally,this ruta y requirements.Wg applie addition, these calculations calculated to be about one rem). All of cask use by any power ructor licensee, conservatively ect the shielding ring the NRC analyses collectively yield h within the United Statea. nerefore. it is which would er reduce dose rate. singular conclusion that & not dependent on any one individusi

60. Comment. Nmenters expressed environmentalimpacts and risks are State's actions including preparation of j

concern that VSC-24 casks were being expected to be extremely amall. a separate DS by any Sute. Further, built at the Palisades Nuclear Plant The absence of significant nothing in this ruta-hg would j before approval or certincation. environmentalimpacts from dry cask oreclude any State from implementing l i I 1 i

17964 raderal Register / Vol. 58, No. 85 / Wednesday, April 7.1993 / Rules and Regulations its envimamental statutes and NRC. De NRC's inspection program to the pmvisions of to GR part 72, regulations as mey otherwise be inckdes reqmresnents to inspect these which includes in subpart K.the permitted by law. gi,sdww. regulations relevant to the storage of

62. Comment. Commenters believed

65. Comment. Commenters stated that spent fuel under a generallicense, A that a cost /banefit analysis abould be the VSC-24 is not a cask. De doofgner spedne referencs to each regulation prepared. One commenter proposed a called it a cask system ^

section is, therefore, uncoosemary cost comparison formula which would Response.De NRC considers it to be

68. Comment. One commenter was estimate costs associated with dry cask e cask. It is called a msk system because favorable to the VSC-24 cask stating "storege over the next 1000 yurs.

it consists of several components. that it was cost. effective, made in the response. A - latory analysis,

66. Comment. Commenters believe U.S.A., additional shielding could be which considers benefits and that there is poor management et added at low cost if required,b impacts of adding the VSC-24 ask to Consumers Power Company, NRC welded closure requires no monitoring, the list of NRC-spproved casks under to Information Notics 91-66 eays they still and risk is minimized by weld sealing CFit part 72, subpart K. was prepared in have a provisionallicense aRar 20 years. the MSB in the reactor fuel handling support of % rulemaking action. It was Consumers Power Company had serious building. Another commenter noted that included as a part af the notice of quality control violations, below average this rulemaking is a pesitive action proposed rulernakiry and is also operating capacity, and faulty which should decrease cost and included in this fmal rulemsking notics. construction at Midland.

incrase h safety of storing fuel This regulatory analysis reflects the R**Ponse. Although this iamment is Another commenter noted b Palisades limited economic scope of this not directly related to this rut mahg, spent fuel poolis closer to Lake rulemakag. The 1000 year cost which is to provide generic approval of Michigan than the cask ped,bo h in comparison identined above assumes the VSC-24 cask design that is not terms of distance and elevation.The 1000. year interim storage at Palisades, dependent on site spec 15e consideration storage of spent fuelin a pool requires an assumption the NRC is not proposing for any one licensee NRC notes that its active systems for shielding cooling and or adopting in this rulemaking.no Systematic Assessment of Licensee reactivity control. %e VSC is passive, NRC Waste Confidence decisions Performancs (SALP) program is an requiring no pumps, valves, or heat concluded there is reasonable assurance integmted staff effort to collect available achangers. the Federal govemment will begin observations and dsts on a periodic Response. None required. receiving spent fuel for disposal by basis and to evaluate licensee es. Comment. Commenters believe:i 2025. Thas, the likehhood of 1000-year Performance, including Consumers that it is not acceptable to increase the interim storese at Palisades is extremely Power, on the basis of thh *nformation. number of approved cask designs. ne small. The most recent SALP report for goal must be the function of the cask

63. Comment. One commenter wanted Palisades, covering the period January 1, itself to contain radioactivity in high letter reports to the NRC distributed to 1991 through March 31,1992, states in concentrations and prevent it from local and state govemment authorities summary, " Overs!! performance at the dispersing into the biosphere as well as and locallibranes in the vicinity of Palisades Nuclear Power Plant was to shield workers and others from facilities using b VSC-24 cask.

characterized by generally steedy or radiation erposure. Some suggested that Response. The NRC laterprets this improving results and showed a altemative actions to dry cask storsgo comment as applying to letter reports conservative and safe operating should be musidered. required by the Cert 15cate of. philosophy.He overall degree of Response. De NRC,in implementing Compliance. Letter reports sent to the management attention and effectiveness the Nuclear Waste Policy Act of 1952 NRC are routinely placed in the Public was acczptable in all areas.** Finally, the has an obligation to approve the use of Document Room and local Public Palisades Nuclear Plant was granted a casks for the storege of rpent fuel. Document Rooms near each facility. full term operating licznse on February provided these casks meet applicable Local Public Document Rooms are 21,1991, regulatory requirements.ne NRC located in public, university, and The SALP report for the proceding a6rees with the commenter that these speciallibraries. A directory of Local period from September 1,1989 through casks should contain radioactivity and Public Document Rooms is published by December 31; 1990 provided similar Protect workers, the public, and the b NRC as NUREG BR-88. he NRC conclusions and stated,"the degree of environment. The previous rulemaking would respond to State requests for management attention and effectiveness of1990 (55 FR 29181) found that spent copies of such reports through NRC's ranged from commendable in some fuel stored in dry storego casks designed State Relations Program areas to needing ettention in others to meet the NRC regulatory

64. Comment. Commenters indicated OveraH. the conduct of activities was ments can mntain radioactivity that operating procedures, evaluation a repriately directed to assurance of ly,his rulemaWg adds one cask reports, and training programs should sa

. Management appeared proactive design, which meets the safety be submitted to the NRC, state and local and ffective in demonstrating a requirements previously developed. De govemment authorities, and placed in conservative operating philosophy and previous see to comments, as local libraries near such facilities. establishing high standards of well as the led safety and Response.These documents expand performance in operations, environmental analyses underlying this on generically approved procedures in maintenance / surveillance, and rulemaking, and described elsewhere in the SAR, CertiScate of Compliancs, or securi." this notice, au reveal that the VSC-24 in the case of the boron determination, 67. ment. One commenter cask will conform to the NRC on national standards. In acxxirdance believed that the CertiScots of requirements, and that its use should with the NRC requirements, liceneses Compliance should list all NRC not pose the potential for significant are not requaed to submit this regulations controlling the use of the environmentalimpacts. information to the NRC or other VSC-24 cask for the storage of spent The principal arternatives available to govemment authorities. Rather, this fuel. the NRC would be procedural in nature, information is evaluated by the licanese Response. no CertiScate of whereby dry cask spent fuel storege and is available for inspection by the Compliance contains a general reference could be opproved under other existing

~ Federal Kagister / Vol. 58. No. 65 I Wednesday. April 7,1993 / Rules and Regulations 17965 or new parts of title 10. Code of Federal review and independent evaluation of radiation sourcs term for other fuels ht Regulations. Regardless of the method the applimat's safety evaluation report may be stored in the VSG-24 cask.NRC selected to approve such dry cask spent and through this rul

action, regulations prohibit Consumers Power fuel storage, all would have similar NRC will assure that h meets part troen using h VSG-24 cask in violation envimnmentalimpacts.

72 requirements and can be used by of the Ceruncate of Compliancs spent The NWPA directed that the NRC individual nue. lear power plant fuel spedfications, and Consumers approve one or more technologies. that licensees with full assuranca of Power must perform written evaluations have been developed and demonstrated protection of the public health and before using the cask that verify all by DOE, for the use of spent fuel storage safety and the enviranment. The NRC Cert 1 Sate of Compliance conditions are at b sites of dvilian nuclear power has experienced no di!Sculty obtaining met. ructors without, to the extent safety information or answers to its As is evident fmm this and ohr practicable, the need for additional site-questions fmm either Erm, either before, responses to public comments, this speciSc review. no NWPA also or aner the divestiture. rulemak f ng provides NRC spproval for directed that the NRC by rulemaking. Following the divestiture.PadBc storage of spent blin the VSG-24 at set forth procedures for licensing the he. lear sent a letter co=taining any site in amordance with the generic technology. Regulations for comments on the VSC-24 design. The conditions and specifications in the accomphshing this are in place. staff natisfactorily resolved and Certiacate of Compliance. As noted,it nerefore, the no action alternative is answered these enmments with a letter; does not constitute a site-specific not acceptable. both the Pacific Nuclear and NRC letters approval of the VSC-24 cask for use by Alternative spent fuel storage are available in the Public Document th" mars Power at the Palisades plant. technologies exist. However, at this Room. no issues contained in this

72. Comment. A number of time, the NRC considers them neither exchange of letters and all other safety commenters requested that the comment sufficiently demonstrated nor issues related to the design of the VSG-period be extended principally citing practicable for use under the general 24 are described in the staffs SER.

the fact that NRC had released a large limnse provisions of subpart K of 10

71. Comment. A commenter noted volume of highly technical material CFR part 72 without additional site-that Consumers Power's comments to associated with the VSC-24 cask and specific reviews. If other storage the NRC during this rulemaking indicate that the 30 day reopening of the technologies become more amenable to that they do not have b kind of fuel comment period which NRC had this type of action.they could be that was spedned in the Certiacate of provided was not a sufficient time for considered at a later time.

Compliance for the casks at Palisades. review and comment on the material.

70. Comment. Commenters expressed ney noted it is hard to believe that the Commentars also questioned why the concern that PaciSc Nuclear. Inc.. the NRC does not know what kind of fuel information was not released earlier.

original designer and manufacturer of it is licensing the cask for, but noted Response.NRCis not granting an the VSC-24 cask system. had ended its that appeared to be the case, ne additional axtension to the comment involvemr st with the cask. Raasons commenter further noted that any period. First, the new information that dted included the issue of liability, approval given by the NRC would have was released is only an increment to negligena issues that might surism in to be site specine and not generic and that previously disclosed. In addition, the future with the cask, the fact that the therefore this would reouire a hearing. most of the individual pages released original designers divested themselves Response.De type of fuel that is are computer output printouts.the due to conmm about the cask. and who being approved for storage in the VSC-results of which were previously would be responsible in the event of 24 cask is speciBed in the vendor's available in various documents made Isakege. Commenters also questioned Safety Analysis Report as well as in the available at the beginning of the public whether NRC had attempted to ucertain Certi5cate of Compliance and SER comment period. In the Federal Register the reason for the divestiture action by prepared by the NRC staff. NRC Notim Ganuary 21.1993; 58 FR 5301) PaciSc Nuclear to discover if le reason regulations require the vendor to specify announcing the comment period related to safety of the cask, liability.or the type of spent fuel to be stored in the extension. NRC made clear the limited, any other consequences, cask before NRC approval, and NRC incremental character of the technical flesponse. NRC is not aware of any thoroughly reviewed the vendor's SAR information.no information of the cask sa fety, negligence, liability or legal and spent fuel specl$ cations and made vendor being disclosed at this time concerns which prompted Pacific them appropriate items for public added detail to the information NRC Nuclear. Incorporated to divest itself comment in this rulemaking. previously placed in the Public from the VSC-24 cask. He key Commenters are therefore mistaken in Document Room at the outset of this individual involved in the design and saying the type of fuel to be stored in rulemaking. It complements and development of the VSC-24 wu also the VSC-24 cask is not known. supplements the design informatien involved in the design and development ne kind of fuel to be loaded into and already dMM providing further of a new modular horizontal concrete stored in the VSC-24 cask at Palisades. detail on such matters as the vendor's spent fuel storage system (NUHOMS should Consumers Power proceed with design calculations (often in the form of design) and formed a new company. use of the VSC-24 cask, must be mmputer runs) and specific data inputs Pacific Sierra Nuclear, for the ecceptable fuel for storage in the cask for models used by the vendor for such commercial manufacture and marketing and must meet the Certificate of calculations, as well as cask design of the VSC-24 storeSe system.NRC Comp!Iance speciScations mentioned details such as reinforcing steel sizf r g focuses its efforts on assuring safety and above for acr:eptable fuel which may be and shield lid thickness.He environmental protection through stored in the cask. In this regard the information being disclosed therefore reviews of apphcations for licenses and Certi5cate of Compliance and SER have provides additional spedficity for the safety analysis reports. lie new been clarined to speci5cally identify the public abo 2t the technical information company spplies for a certi5cate of fuel assembly classes acceptable for that was considered by the NRC staff an compliance, that new company must storage in the VSC-24 cask and to pre ng the principal NRC doc.nr.ts meet all NRC requirements as would identify limits for physical dimensions, un rlying this rulemaking.The se any existing mmpany. nrough NRC's weight, burnup. decay power, and documents include the proposed

17964 Federal Register / Vol. 58, No. 65 / Wednesday April 7,1993 / Rules and Regulations Certi$cate of Comphanc= is the cask

75. r^==aa' Namantars asked who impact on which this ds4ermination is and the associated NRC staf SER and would be r=ada for oversight of beesd is available for is7w t the a

related EA, which w are previously bl stcred in casks after NRC Public Document Room. 21201, placed in the NRC Public Docurnent decommi=*w f b roedor. Street, NW. (lower levell, Washington. o Room at the outset of this propcsed shipment of the bel off-site,and for DC. Single copies of the Environmental rulemaking. decommissioning of the casks aRar Asseeement and the Finding of No Second, the initial public corr. ment stored fuel was ahlpped off-site. Signi5 cant Impact see evailable from peried opened on June 26.1992, and Response. In accordance with 10 CFR Mr. Cordon E. Gunderson, Office of closed on September 9,1992. The 50.54(bbl. all operating nuclear power Nuclear Reg 41 story Research. U.S. comment period was reopened on reactor hcensees are required, no later Nuclear Regulatory ranmission, January 21.1993 and ended on February than 5 years prior to the expiration of Washington, DC 20555, telephone (301) 22,1993,in addition, at the pubuc the openting licaces to provide the 492-3803. meeting held with the Michigan NRC, fc'r review and approval, the Attorney General on February 23,1993, licenese a program to = nap and Paperwork Redoction Act Stan===ar NRC assured that comments received pmvide funding for the management of This Anal rule does not contain a new within five working days aAer that all irradiated bl. NRC's review of the or amended information colledion meeting would be considered. Ahhough limnaee's bl management pmgram requimment sub)ed to the Paperwork the comment periods have closed.NRC will be undataken as part of cactinued Reduction Act of1980 (44 U.S.C. 3501 has considemd all comments received. lir=nemg under the pmvisions of part 50 et esq.). Existing reguirements were Thus,the pubhe comment period for and part 72 of the =mi= ion's appmvod by the Omcm of Management r this rule has effectively been almost r lations. and Budget approval number 3150-nine months which the NRC believes ith resped to riammmi-wing, the 0132. constitutes more than suf$cient time for licensee may select a d-=1-wing Reguletary Analyses altem ve that wil this type of rulemsktng

73. Comment. One emnmentn spent fuel pool in whic case the Commisalon issued an amendment to 10 questioned the validity of neglecting licensee will be' trquired to maintain its CFR part 72. which provided for the Eunma does at the nozales.

8 "*' D "I"'"C*d Can 5 part 50 license: storage of spent nuclear bl under a calcufatos the does rate as the MSB is

2. Allow storage of fuelin a carti6ed pneral11ran= Any nuclear power lowered into the VCC during transfer.

cask under the provisions of part 72 as reactor licensee can use these casks if: long as the part 50 bconse r=maina in (1)hy notify the NRC in advance;(2) Does is estimated at the point of effect; or the spent fuelis stored under the maximum exposure, that 13. at the outlet

3. Allow storsgo in an on-site cxmditions spec 25ed in the cask's vent and the top of the VSC. Under independent spent bl storage Ceniha or Compliance; and (3) the these circumstances, the entire instauntion under the alts-spedSc other conditions of the generallicense distnbution of radioactive material in licensing provisions ot part 72.

am mt. As part of the 1990 rulemaking, the t bl assemblies contributes to For any of the above alternatives, the four spent fuel storey casks were the me in a transient fashion.no Hesnses wiu be responsible for safe approved for use at res.ctor sitas, and assumation that the source is directlY stange of spent bl during the period were listed in to CFR 72.214. ht from tEs active fuel which is aliped of storage, for later shipment off-site for rut =,nmag envialoned that storage with the air exhaust is conservstive, further storage or disposal and for anal casks carti5ed in the future could be since it is the highest and is sustained decommissioning of the reactor routinely added to the listing in $ 72.214 for a short period of time. Other MSB/ fuel pool. dry storage caak or to through rulemutng procedures. VCC relative positions during transfer a level permitting unrestrided release of Procedures and criteria for obtaining would yield smaller done rates. the site and facihty.h requirements NRC approval of new spent fuel storage Calculations demonstrated that the does for decomm8=wian are provided in to czak designs were provided in 10 CFR rate from gamme-emitting radioactiv* CFR part 72.30, which defines 72.230. materialin the nozzle is three orders of decommi=4aaw planning, fanr< 1 ne ahernative in this proposed magnitude less than the does rate fran assurance and roccedkeeping action is to withhn1<1 certi$ cation of the active fuel section. provisioca. these new designs and to consider the

74. Comment. A commenter noted granting of a site-speciHe license to each that the geornetry for does calculations Findlag of NWgar=*

utility that applied for permission to use was based on an earlier desip and not Envirn====tal hupact: AveBobGitF these new caska. His alternative would on the latest conEguration. Under the NationalEnvironmental be mars costly and time consuming Rerponse. N chanys la design Policy Ad of1969, as amended, and the because each site-speciSc license referred to by the enmmenter were all ht enmmission's regulations in subpert A appliantion would require a specific S repositionings of the inlet air duct. N of 10 CFR 51, the %==I= ton has review. In addition, withholding r reorientation involves minor changes of determi that this rule is not a major certiKration would ignors the both the honzontal and vertical Federal action signiRran'Sy a5ecting the rulemaking procmdures and criterie in orientation of the duct but does not quality of the human environment and, to CFR part 72, subparts K and 1. for change the circuitous path which thersfors, an environmentalim the addirlan of new cask designa, contributes to radiation protection. In statement is not required.nis ' rule Further. It is in conflict with the addition, the analysis does not take adds an additional cask to the hat of Congressional direction in sect 3ons 133 credit for the 0.5-inch stealliner of the approved spent bl storage casks that and 218 of the Nuclear Waste Pohey Ad duct which would offset any small power reactor licensees can use to stars of 1962 to establish procsdures for the changes in dose due to roodentation of spent fuel at reactor sites without licensing of *=cha^1of es for the use of f the duct. brefore, the design changes additional site-specine approvals by the spent bl storage at the sites of cmlian do not result in a signincant change in Commi= ion. N environmental nuclear power ranetars without, to the the radiation dose rate calculations. assessment and finding of no signi5 cant extent pr=r+tenhie, the need for

A I ~ Tederal Ragister / Vol. 58. No. 65 / Wednesdayo April 7.1993 / Rules and Raydah= 170s7 j additional site reviews. Also, thia issued by the Small Bustaana 872.214 ust at sevseeed spans toes i alternative would axclude new vendor Administration at 13 CFR part 121. storego esena. cask designs from the approved NRC hst under subpart K without cause and h@ Cseba= Numbac 1987 ) would arbitranly limit choice of cask no NRC has determined that the SAR Submitted by: PactSc $5erra Nuclear desips avs.ilable to power reactor beckEt rule to CFR 50.109. does not Asmodetse bcensees under the general hcense. 8Fply to this Enal rule, and thus, a SAR

Title:

Seis'y Ana?yris Report Ibr the This Ensi rulemaking will ehminate backat anah sis is not required for this ventileted Samass Cask System j k he 72-1 the above problems. Further. this action Enal rule, becaum this amendment doce i Gnu will have no adverse effect on the pubhc DM involve EDY Provisions which would W 7 *3- %gm beelth and safety. impose backEts as deEned in The beneSt of this Enal rule to $ 50.109(a)(1). DeedW W W e d g y119s3. nuclear power ructor licensees is to List of Subjects in to CFR Part 72 Far the Nuclear Regulatory Chumlerkm. reske avallable a ter choice of spent Manpower training .Nuclur leaseH.Sansamk. materials.Oxupationb fuel stwage cask es wM can be and Acting f.recuche Directorfbr operwoom. r casb ps m*a y health. Reporting and recordkeeping trR Doc. 93-st12 Filed 4-4.e3; 3.45 aml n have an advantage over the existing [ments. Securuy measures. Spent usses ooon meew designs in that power reacts licensees For the rusons set out ir the may or may not prefer to use the newe' preamble and under the tesrity of the casks. ne new cask vendors Mth casks Atomic Energy Act of 1954, as amended, to be usted in 5 72.214 beoeSt by being the Energy Reorganization Act of1974, able to obtain NRC cert 1Ecstas once for as amended, and 5 U.SE $52 and 553, a cask desip which can then be used the NRC is adopting the following by many power reactor licensees under amendments to to CFR part 72. the general license. Vendors with czak desips altsady listed may be adversely PART 72. UCENSING impacted in that power reactor licensees REQUIREMENTS FOR THE may choose a nowly listed design over INDEPENDENT STORAGE OF SPENT an existing one. However, the NRC is NUCLEAR FUEL AND HIGH LIVEL required by its rvgulations and NWPA RAD 60 ACTIVE WASTE requirements to estabush a procedure and to consider appucations to certify 1.He authority citation for part 72 and list approved caska. no NRCalso continues to read as follows: beno6ts because it will be able to certify Authorirr Sees. St. $3. SF. 82. 43,65,69. a cask desip based on one generic st.1at.1s2.1as. te4,1sa.1s7.188 as Stat. safety and environmental review. for s2s. s30. an. 533 s34. s35,948, tE2, est. use by multiple licensees.His Enal 955, as amended, sec. m. s3 Stet 444. es amended (42 USC 2071. 2073. 2077. 2W2. rulemaking hu no si **'Ecant 6 2093.2005.2099.2111,2201,2232.1233 identi6sble impact or beneEt on other 2m. 2236. 2237. 2134. 22421: sec. 274. PA government agences. L 86-373,73 Stat. 6&s. as amanded (42 Based on the above discussian of the U.SC 2021): sec. 201, as a:nonded. 202. 206. bene $ts and impacts of the alternatives, &s Stat.1242, es== dad 1244,1248 (42 the NRC concludes that the USC. 5841. 5442, sase): Pub. L es-act. sec. requirements of the anal rule are

10. 92 Stat 2est (42 U.SC sa51k sec ta2 commensurate with the Commission's Pub. L 9t-193. 83 Stat $53 (42 UM 4332k escs. m.132. m.135.137. m. M L 97-responsibilities for protection of the 42S. 96 Stat. 2229,2230,2232,2241. sec. t48, public bealth and ufety and the PA L te. tot Sat mo-ns H2 comman defense and ucurity.No other U.SC totst, toss 2. tote 3 tot $s.totsf.

available alternative is beUeved to be as 10181, toise). satisfactory; thus, this action is h* 72.44(g) also lesued under esca, recommended. 142(b) and t es(c),(dL Puh. L 20c+203, tot Stat.1330-132,1330-234 (42 U.SC Regulatwy UzzibMy Act CertiEcarica tots 2(bL wisa(cXdit Section 72.4s slao In accordance with the R*6ulatory issued under sec.1st. 6s Stat. 965 (42 UE Flexibility Act. 5 U.SE 6056'b), the 2mL sec. m. PA L 87-425. 98 $mt mo Commission certines that this rule, will B2 ESC tomt Section 723s(d)also issued under sec.145(g). Pub. L 100-202, not have a signiEcant economic impact tot Stet 1330-235 (42 UE totte(s)L on a substantial number of small Subpart I also issued under sees 2tzk 2(15). entities. His amendment affects only 2(1e).117(at 141[h). Puh. L 97-425,98 Stat, licensees owniag and operating nuclear 22cz. 2203. 2204. 2222. 2244 (42 Um power reactors and cask vendors.The 10101. 2ot37(a), totst(h)L Subperts K and L owners of nuclear power plant 5 do not are also issued under sec.133. to Stet 2230 fall within the sccpe of the definition of (42 UE 20153) and sec. 218(ab 96 Stat. "small entities" set forth in set tion 2252 H2 M W 9sk 6C1(3) of the Regulatory Flexibility Act,

2. In $ 72.214. CertiEcate of 15 U.SC 632, or the Small Businees Coccpliance 1007 is added to read as Size Standards est out in regulations foUows:

l-w; am EDO Principal Correspondence Control FROM: DUE: 10/29/93 EDO CONTROL:.0009447. DOC DT:-10/06/93 FINAL REPLY: -t JAMES KLEINHANS, CHAIRMAN WISCONSIN RADIOACTIVE WASTE REVIEW BOARD TO: SECY =FOR SIGNATURE OF :

• • GRN CRC NO: 93-0933

-MURLEY DESC: ROUTING: .i SPENT FUEL REMOVAL FROM POINT BEACH TAYLOR SNIEZEK THOMPSON BLAHA BERNERO,'NMSS SCINTO,- OGC DATE: 10/21/93 ROTHSCHILD, OGC r ASSIGNED TO: CONTACT:

i NRR MURLEY SPECIAL INSTRUCTIONS OR REMARKS:

i OGC CONTACT IS TRIP ROTHSCHILD. NRR RECEIVED: OCTOBER 21, 1993 NRR ACTION: DRPW: ROE NRR ROUTING TEM /FJM JP WR DC I FG I KB I I l

), '. OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL-TICKET PAPER NUMBER: CRC-93-0933 LOGGING DATE: Oct'21=93 I i ACTION OFFICE: EDO i AUTHOR: JAMES KLEINHANS AFFILIATION: WI (WISCONSIN) ADDRESSEE: SECRETARY LETTER DATE: Oct 6 93 FILE CODE: IDR-5-POINT BEACH

SUBJECT:

SPENT FUEL REMOVAL FROM POINT BEACH ACTION: Direct Reply DISTRIBUTION: CHAIRMAN, COMRS, OGC, CAA, OCA,OPA SPECIAL HANDLING: COORD WITH OGC CONSTITUENT: i NOTES: OGC CONTACT:-TRIP ROTHSCHILD i DATE DUE: Oct 29 93 ) SIGNATURE: DATE SIGNED: AFFILIATION: l' i i J .s e i-i i EDO --- 009447 'l ^^ ,}}