ML20217P283

Forwards Copy of Petition for Rulemaking (PRM-72-4) Requesting That NRC Amend 10CFR72 Regulations That Govern Independent Storage of Spent Nuclear Fuel in Dry Storage Casks

ML20217P283
Person / Time
Issue date:	03/10/1998
From:	Rathbun D NRC OFFICE OF CONGRESSIONAL AFFAIRS (OCA)
To:	Inhofe J, Schaefer D HOUSE OF REP., ENERGY & COMMERCE, SENATE, ENVIRONMENT & PUBLIC WORKS
References
NUDOCS 9804100055
Download: ML20217P283 (2)

Text

{{#Wiki_filter:i i pm atg ' / ? ' t:bd_ g 'e UNITED STATES g ,j NUCLEAR REGULATORY COMMISSION f WASHINGTON, D.C. 30666 4001 e%,,,,,/ March 10, 1998 i The Honorable James M. Inhofe, Chairman i Subcommittee on Clean Air, Wetlands, Private Property and Nuclear Safety Committee on Public Works United States Senate Washington, DC 20510

Dear Mr. Chairman:

J Enclosed for your information is a copy of a petition for rulemaking (PRM-72-4) requesting that the Nuclear Regulatory Commission (NRC) amend 10 CFR Part 72. The petition was filed with the NRC by the Prairie Island Coalition. l The petition requests that the NRC amend its regulations that govem independent storage of spent nuclear fuel in dry storage casks. Also enclosed is a copy of the Federal Register notice that contains additional information conceming the petition. The notice will be published requesting public comment for a 75-day period. Sincerely, i j Dennis K. Rathbun, Director Office of Congressional Affairs

Enclosures:

1. PRM-72-4

2. Federal Register notice cc: Senator Bob Graham l

l \\j 9804100055 980310 PDR ORG NRCCO ('v PDR

,p* **h '4 UNITED STATES s j NUCLEAR REGULATORY COMMISSION 2 WASHINGTON, D.C. 20066-0001 %,y,, + March 10, 1998 The Honorable Dan Schaefer, Chairman Subcommittee on Energy and Power Committee on Commerce United States House of Representatives Washington, DC 20515

Dear Mr. Chairman:

Enclosed for your information is a copy of a petition for rulemaking (PRM-72-4) requesting that the Nuclear Regulatory Commission (NRC) amend 10 CFR Part 72. The petition was filed with the NRC by the Prairie Island Coalition. The petition requests that the NRC amend its regulations that govem independent storage of spent nuclear fuel in dry storage casks. Also enclosed is a copy of the Federal Register notice that contains additional information concoming the petition. The notice will be published requesting public comment for a 75-day period. Sincerely, Dennis K. Rathbun, Director Office of Congressional Affairs

Enclosures:

1. PRM-72-4

2. Federal Register notice cc: Representative Ralph Hall 1

l l

[ l ( 1. 2 k [ L _1 L _aL P.O. Box 174 - Lake Elmo. MN 55042 - Phone: 612 770 3861 FAX 770-3976 August 26 l# L Joseph Callan Executive Director of Operations US Nuclear Regulatory Commission Washington. D C 20555 Dear Ntr Callan Please find enclosed a petition pursuant to Section : 200. Title 10 of the Code of Federal Regulations (CFR) The Prairie Island Coaktion < Plc i hereby petitions the Nuciear

-e:._ n :.:y Commission (NRCi to suspend for cause the Northern States Power Co MPi Niaterials License No SNN1-250o oeeded to operate an Independent Spent Fuel Storage Installation (ISFSI) at the Prairie Island Nuclear Generating Plant O'l)

A thorough resiew of the procedure deseloped by NSP for unloading Transnuclear drs sterage casks (TN-40)in use at Pl is necessarv at this time because it is apparent that conditions for safel) unloading TN-40 casks after a storage penod hase not been established By operating the ISFSI at PI prior to estabhshing safe unloading conditions. NSP is violating the requirements of 10 CFR 72122(lI and other rules and regulations of the United States Nuclear Regulatorv Commission Toward this end. Petitioner also requests formal rulemaking proceedings under 5 U S C 553(e)to examine the issues addressed herein Thank you Sincerely, P d Mf George Crocker. Steering Committee Prairie Island Coalition ~ c: ,I ')

.o BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION Docket 72-10 IN THE MATTER OF: ) ) PRAIRIE ISLAND COALITION, ) Petitioner, ) ) ) ) ) ) vs. ) UNITED STATES NUCLEAR ) REGULATORY COMMISSION, ) Respondent. ) Petition Pursuant to 10 CFR Part 2.206 of the Commission's regulations, the Prairie Island Coalition petitions the Nuclear Regulatory Commission (NRC) to: l. Suspend Northern States Power Co.'s (hereinafter "NSP") Materials License No. SNM-2506 for cause under 10 CFR 50.100 until all material issues regarding the maintenance, unloading, and decommissioning processes and procedures, as described in this Petition of the Prairie Island Coalition, and also that of the Prairie Island Indian Community's recent S2.206 Petition, incorporated herein by reference, have been adequately addressed and resolved, and until the maintenance and unloading processes and procedures in question are safely demonstrated under the scrutiny of independent third party review of the TN-40 cask seal maintenance and unloading procedure. 2. Determine that NSP violated 10 CFR 572.122 (f) by using a cask design that requires periodic seal maintenance and emergency seal replacement that must be performed in the plant storage pool. But these casks cannot be placed back into the pool to perform these functions due to unresolved problems with fuel degradation during storage, flash steam, thermal shock, and the resulting potential for radiation dispersion..,,ga,.g -.Sp (-l,.

I l .' o 1 1 3. Determine that NSP violated 10 CFR 572.122 (h) by using a J . cask that must be placed into the pool for necessary maintenance and/or unloading procedures, while such placement will prematurely degrade the fuel and pose operational safety problems with respect to its ultimate and necessary removal from dry cask storage. 4. Determine that NSP violated 10 CFR 572.122 (1) by loading casks and storing them under their license before it had l procedures adequate to safely unload and decommission the l TN-40 casks. 1 l l 5. Determine that NSP violated 10 CFR 572.130 by using the TN-l 40 cask and failing to make provisions capable of l accomplishing the removal of radioactive wastes and contaminated materials at the time the ISFSI is permanently i decommissioned. This failure may prevent decommissioning. 6. Determine that NSP violated 10 CTR 572.11 by failing to provide and include complete and accurate material information regarding maintenance and unloading of TN-40 casks in their ISFSI application and in subsequent submissions regarding cask maintenance and unloading issues. 7 Determine that NSP violated 10 CFR 572.12 by delib*erately and knowingly submitting incomplete and inaccurate material information regarding maintenance and unloading of TN-40 casks in their ISFSI application and in subsequent ' submissions regarding cask maintenance and unloading issues. 8. Require that NSP pay a substantial penalty for each cask that the utility has loaded in violation of NRC regulations. 9. Administer such other sanctions for the above violations of NRC regulations as the NRC deems necessary and appropriate. 10. Provide Petitioner the opportunity to participate in a i public review of maintenance, unloading, and decommissioning i processes and procedures in question and an opportunity to comment on draft findings after investigation by the NRC. j 11. Order modification of NSP's Technical Specifications to ensure a demonstrated ability to in fact safely maintain, unload, and decommission TN-40 casks. 12. Review NSP's processes and procedures for maintenance, unloading, and decommissioning, and if NSP does not possess capability to unload casks, order NSP to build a " Hot Shop" for air unloading of casks and transfer of the fuel. 13. Under 5 U.S.C. 553 (e), Petitioner requests a formal . rulemaking proceeding to solicit information and review current information regarding thermal shock and corrosion inherent in dry cask storage and usage and to define the parameters of degradation acceptable under 10 CFR 72.122(h). 14. Under 5 U.S.C. 553 (e), Petitioner requests a formal rulemaking proceeding to define the parameters of retrievability required under 10 CFR 72.122 (1). 15. Under 5 U.S.C. 553(e), Petitioner requests a formal rulemaking proceeding for amendment of current licenses and l rules for prospective licensing proceedings to require demonstration of a safe cask unloading ability before a cask may be used at an ISFSI.

• Preliminary Matters and Facts 1.

The Prairie Island Coalition (he'reinafter "PIC") incorporates herein by reference the facts, argument, and l conclusions of the Prairie Island Indian Community's 52.206 Petition dated May 28, 1997. 2. PIC was established in 1990 for the purpose of location and dissemination of information regarding dry cask storage, and+ opposition to NSP's plans to construct and operate an Independent Spent Fuel Storage Installation (hereigafter "IS FS I" ) at its Prairie Island Nuclear Generating Station (hereinafter "PI"). PIC is a coalition of 30 environmental groups, tribal and urban Indian organizations, peace and justice groups, businesses, religious groups, and urban and rural citizen organizations. It is a project of the North American Water Office. 3. At the state level, PIC has been actively involved in Minnesota public decision-making proceedings regarding PI nuclear generation and nuclear waste. This involvement includes formal intervention in the " Certificate of Need" proceeding before the Minnesota Public Utilities Commission, litigation.in state courts regarding the Certificate of Need, and on-going legislative and educational efforts on nuclear waste and nuclear generation issues. '4. At the federal level, PIC has an active relationship with the NRC regarding PI nuclear operations. PIC filed a 52.206 petition with the NRC on June 5, 1995 regarding failure of reactor components and waste management problems, including cask unloading problems. PIC participated in the NRC public meeting in Red Wing, MN regarding NSP and Transnuclear TN-40 cask fabrication quality control problems. PIC petitioned for intervenor status in NSP's licensing proceeding before 't l the NRC regarding a site in Florence Township for an . alternate site to store nuclear waste. PIC has also monitored NRC meetings in Washington, D.C., regarding waste issues, and has met and exchanged written communications with NRC staff about these issues. 5. In a February 25, 1997 letter from Gail H. Marcus of the NRC staff, Ms. Marcus acknowledged that there is no "... actual experience in unloading spent fuel from a cask following a I long period of storage..." Exhibit A, February 25, 1997 l Letter from NRC's Gail Marcus to George Crocker, Steering l Committee, Prairie Island Coalition. Ms. Marcus states that instead, the NRC staff rely on a " general understanding" of technical capabilities and related experiences to assess the adequacy of a licensee's' procedures for unloading dry storage casks that have contained irradiated fuel for a i period of time. 6. Irradiated fuel in storage casks'will experience thermal shock when a cask is reflooded prior to unloading. Thermal shock may degrade fuel assemblies, perhaps extremely dramatically. Degraded fuel assemblies can increase radiation exposure to workers and off-site due to the compounded difficulty of adequately isolating irradiated l fuel debris, the increased venting of radioactive gasses from the increased number of fissures in the debris, and the potential involvement of criticality issues. In thq February 25, 1997 letter, Ms. Marcus recognizes that "...the limited unloading experiences with storage casks have not involved temperature differences between fuel and coolant..." and that such differences create the potential for " thermal shocking." There have been no procedures developed to protect operation safety if thermal shocking occurs, and no assessment of how those procedures impact worker or off-site radiation exposure. 7. Thermal shock may cause fuel assembly degradation. In the February 25, 1997 latter, Ms. Marcus acknowledges that fuel disintegration patterns could lead to fuel reactivity for criticality considerations. She states that, "Upon detection that fuel disintegration has occurred, special measures would be developed and implemented to assure an adequate safety margin is maintained during unloading." In i other words, the measures have not been developed, and there has been no assessment or evaluation regarding the actual ability of such measures to adequately protect worker and pubic health, and the environment. Safety margin references may also be assumed to refer to the question of whether the disintegrated fuel could be physically unloaded at all. 1 i A O 8. Also in this letter, Ms. Marcus reaffirms that SARs "over- . simplify matters" when they state that unloading is basically the reverse of loading, because such statements do not reflect that the unloading process introduces different conditions and complications compared to the unloading process. l 9. In a letter dated July 10, 1997, from Beth A. Wetzel of the l NRC staff to NSP, Ms. Wetzel requests additional information regarding the PI spent fuel special ventilation technical j specifications. Exhibit B, July 10, 1997, Letter from NRC's Beth A. Wetzel to Roger O. Anderson, Director of Licensing and Management Issues for NSP. In this request, Ms. Wetzel has clearly acknowledged the importance of the considerations which she. raises, taking these concerns a l l step further than Ms. Marcus in her letter (Ex. A), 1 particularly regarding concerns about steam pressurization when the cask is initially filled with radioactive pool water prior to loading. This request raises valid questions about the ability of the l pool ventilation system to adequately vent, contain, and l filter radioactive material coming out of the cask as the water enters. Ms. Wetzel acknowledges the potential for thermal shock, and that a cask unloading procedure" which produces this effect may result in significant radioactive contamination of the environment. Degradation of the fuel and/or assemblies due to thermal shock is equally froubling. \\ \\ i l 10. It has long been known that unloading is more complicated and wholly distinct from loading. This fact is confirmed in a study of the unloading of Transnuclear's TN-24P, where over time, the material stored in the cask was misshaped and impossible to remove. Exhibit C, October 18, 1990, INEL Letter from Schmitt to Fischer. Exhibit D, November 21, 1990, INEL Letter from Schmitt to Fischer. j 11. On April 16, 1997, Jack W. Roe of the NRC sent an internal memo to another Staff member defining NRC's dry cask storage l terms. Exhibit E, April 16, 1997 NRC Memo from Jack Roe, Director, Division of Reactor Projects III/IV, Office of Nuclear Reactor Regulation, to Cynthia D. Pederson, Director, Division of Nuclear Materials Safety, Region III. This memorandum offers " clarifications regarding the terms ready retrieval and structural defects." In this memorandum, Mr. Roe defines " ready retrieval" to mean that the regulations do not require licensees to be able to immediately retrieve waste. See 10 C.F.R. 572.122(1). In his explanation of why licensee's ability to " someday, somehow, maybe" retrieve spent fuel from storage would meet the regulatory requirements, he fails to take into account 1 \\, the physical realities, problems and constraints identified .by Ms. Marcus in her letter of February 25, 1997, or the difficulties encountered in the INEL study where the material simply could not be unloaded due to deformities and changes over time. 1 12. Mr. Roe also stated that: [S]taff has not identified the unloading of a cask as a required protective measure to be taken within a specified time in order to limit the offsite consequences of an accident involving the release of radioactive material from a storage cask. Id. This is Mr. Roe's ra'tionale for allowing a utility to operate where there is not enough room in the spent fuel pool to unload immediately, i.e., at Prairie Island, or where a spent fuel cask has weld flaws, i.e., Palisades, where welds have failed. Mr. Rob did not address the issue or assurance that the utility can in fact unload the casks. 13. There are other reasons to unload a cask that have not been addressed in Mr. Roe's letter. The NRC has clearly stated that: (S)hield-lid weld failures affect the integrity of a cask confinement boundary. The root-cause of the shield-lid failures and the potential for dela'yed cracking on loaded casks must be understood. Although the failure of both the cask's inner shield-lid seal weld and outer structural-lid weld would not pose an off-site threat to public health and safety, such an occurrence would cause the loss of the helium atmosphere inside the cask. This loss could result in cladding degradation and future fuel handling and i retrievability problems. Since one of the design requirements of the cask is the.long-term protection of the fuel cladding [10 CFR 122 (h)], such degradation would be unacceptable. Exhibit F, April 15, 1997, Letter of NRC Inspection Report. Mr. Roe's rationale does not address the potential for helium leaks inherent in failed welds that would cause unacceptable degradation. A similar credible event at Prairie Island would be the occurrence of a leak in the cask seals. In such a situation, whether the cask can be unloaded immediately is not the issue. The issue is whether it can, in fact, be unloaded at all. For over two years, Consumers Power has demonstrated that it is unable to unload the cask with failed welds. 't

14..Another reason casks must be placed into the pool and opened is obligatory cask maintenance which must be completed on Transnuclear's TN-40 cask.

Exhibit G, NSP SAR for Prairie Island ISFSI, Table 5.1-2. Seals must be replaced, or again, there will be a helium leak and unacceptable degradation. It also does not address whether NSP can replace a seal on a cask 20 years after it was loaded or when a seal fails. And seals do fail. Again, Mr. Roe's rationale does not address whether the cask can, in fact, be i unloaded. i 15. Another reason the TN-40 casks at Prairie Island would require unloading is that state law requires that they be moved off of Prairie Island. This state requirement anticipates that the casks must be moved after a term of temporary storage, in Minnesota defined as eight years. In the matter of Spent Puel Storage Installation, 501 N.W.2d 638 LMinn. Ct. App. 1993). Even' if the spent fuel were to stay for the life of the NRC license, it would have to be unloaded to move to a federal interim site or repository, as provided in the NRC's Waste Confidence Decision and upon which all nuclear waste storage facilities are premised. September 11, 1990, Waste Confidence Decision Review, 54 CFR 39767. Again, this is another scenario where the'NRC's anticipation of the necessity of unloading is inadequate. 16. Yet another scenario where unloading is required id for decommissioning. NRC authority rests on the requirement that it license only facilities that can be constructed, operated, and decommissioned. NRC regulations require that the facility "be designed for decommissioning," and that the licensee make provisions to " facilitate the removal of radioactive wastes and contaminated materials at the time the ISFSI...is permanently decommissioned." Because there are unaddressed unloading issues such that it is unreasonable to assume that the TN-40 cask can indeed be unloaded, NSP has violated the rule by failing to make the required provisions that assure it can decommission the licensed facility. 17. There is an important distinction to be made between immediate cask unloading and the actual ability to unload a cask. Mr. Roe is correct in that the NRC's rules do not require a licensee be able to immediately unload a cask. The NRC rules do clearly require that a licensee be able to unload a cask. The technical difficulties that have been documented thus far give sufficient reason to doubt a cask can be unloaded in a pool if it has been used for storage for some time. Further, because unloading in a pool has not been completed, there is sufficient reason to require that a 1 l \\ utility demonstrate that it can unload a cask. If the . utilities can demonstrate that a cask can be unloaded in a pool after long-term storage, we can rest assured with the knowledge that,'although they may not be able to unload it as soon as the'need to unload appears, they will in fact be able to unload it at some reasonable point in time. 18. No dry cask that has been used for storage for some time, i.e., over a year, has been unloaded in a pool. There are l issues that remain unaddressed, and NSP has not demonstrated l .that it is able to unload a cask in its pool. It has no other facilities for unloading. 19. The NRC itself declares that cladding degradation, because I it could lead to future fuel handling and retrievability l problems, is unacceptable. Ex. F, 4/15/97 NRC's Susan Frant l Shankman Letter to Sierra Nuclear. In that particular case, I the letter writer is concerned with degradation due to escape of helium, and emphasizes'that: 1 since one of the design requirements of the cask is the long-term protection of the fuel cladding [10 CFR 72.122 (h), such degradation would be unacceptable. Loss of helium from the TN-40 cask is an anticipat*ed event, hence NSP's seal = pressure monitors. Exhibit H, June 30, 1995, Notice of Violation, Inspection Report, 7.1, p. 23. However, the degradation that a helium leak would dause is not addressed, nor is the method by which NSP would replace the defective seal. NSP's TN-40 cask runs the significant risk of degradation due to thermal shock, loss of helium through failed seals, and most importantly, degradation due to the passage of time. NSP's TN-40, its seal maintenance program, thermal shock inherent in placing the cask in the pool, and degradation over the passage of time make this cask unsuitable for storage. NSP is therefore in violation of 10 CFR 72.122 (h). 20. In a study of the TN-24P, which NSP claims is.so very similar to the TN-40, conducted by INEL in 1990, INEL experienced serious thermal problems, not related to cladding, but to the structure of the inserted canisters. Exhibit C, INEL Letter, October 18, 1990; Exhibit D, INEL Letter, November 21, 1990. It is important to note that these were canisters containing assemblies, which allowed less room in the basket. It is equally important to note that these casks were unloaded in air in a Hot Shop. These canisters had been stored for several years, and the thermal damage was so severe that the canister could not be \\ ? unloaded. In the October 18, 1990 letter, the writer . declared: [T]he canisters had " setup" in some fashion: thermally, twisting, bowing, corrosion or other..." The canisters had apparently taken on a set most probably thermally induced although possibly including other factors such as bowing, twisting or other. The laminated makeup of the TN-24 basket may also be involved...It should be clear, nevertheless, that the experience encountered should receive future focus since the inability to extract at lest one of the assemblies with existing equipment is apparent. In the November 21, 1990, letter, in the " Review of Stuck Puel Assembly Issue," the writer said of the damage: [T]hermal expansion of the' canister is the most probable cause, bowing, twisting or other mechanisms cannot be eliminated as possibilities; we presently have little capability to determine the root cause because accessing the assembly.or the basket is not feasible with fuel in the cask. For the other six canisters in the TN-24P, it is possible, alth'ough not probable, that additional canisters may be unremovable, it is also possible that canister number 18 iq no longer stuck because of thermal unloading of the basket following the removal and placement in the VSC-17 cask of 17 fuel canisters. Id. The letter noted that an attempt could be made to remove the stuck canister, but a major consideration was that it "may become stuck in a partially withdrawn position or that canister damage might be incurred." Clearly, fuel stored in the TN-24P is not retrievable. l 21. NSP's SAR for the Prairie Island ISFSI provides that the TN-40 cask seals must be replaced every 20 years, or sooner if there is a seal failure. Exhibit G. The SAR states that as a part of the cask seal replacement, the TN-40 must be i placed in the spent fuel pool, and that replacement of the seals is completed in the pool. Yet, as demonstrated by Beth A. Wetzel's 7/10/97 Request, there are unresolved safety considerations recognized by the NRC, primarily ventilation of the flash steam produced by introduction of the cooler pool water into the hot cask. Exhibit B, 7/10/97, License Amendment, Request to NSP. Secondly, there remain unresolved thermal shock issues, where introduction of cooler pool water would crack zircaloy cladding or assemblies.,

i i l

22..NSP consistently claims that casks can be unloaded, and that

" thousands of Transnuclear casks have been unloaded worldwide." Exhibit I, Environmental News, August 1997. l NSP has also made this statement under oath in an Affidavit, and in its legal argument. Exhibit J, In the Matter of a Request by Northern States Power Company for Certification of Compliance, Cl-96-2189, C8-96-2190, Respondent's Response, p. 5-6; Aff. of Jon Kapitz, p.2. In Mr. Kapitz's Affidavit, he first states that: The unloading: procedure and the relevant design features for taa TN-40 casks approved for use at the PI

• Plant are based upon features and procedures common to existing Transnuclear casks used worldwide, including shipping casks and storage casks like the TN-24P cask.

Exhibit J, Aff. of Kapitz, p. 2 (emphasis added). He goes on to say that: While NSP has not needed to unload any of the five TN-40 casks that have been loaded at the PI plant to date, a comparable Transnuclear storage cask (a TN-24P cask) has been successfully unloaded as part of a project Jointly sponsored by the Electric Power Research Institute and the United States Department of Energy. Id. (emphasis added). Although it is accepted pradtice to-attach to an Affidavit any source used as the basis for that Affidavit, Mr. Kapitz did not do so! Mr. Kapitz did not even specifically cite the study! 23. Mr. Kapitz's statements are false. He claims that the procedures developed for Prairie Island are the same as those for the TN-24P. However, a fundamental element in NSP's unloading procedure is that it is a pool transfer. A quick review of the study provides a reason it may not have been included with Mr. Kapitz' Affidavit. Exhibit K, EPRI, "The TN-24P PWR Spent-Ebel Storage Cask: Testing and Analyses" EPRI NP-5128, April 1987. The cask to cask transfers in this study were completed in a " Hot Shop" and were AIR transfers. These were not pool transfers as are required at Prairie Island. Hot Shop transfer procedures are inapplicable to pool transfers and do not substantiate any claim that NSP can unload a TN-40 in a pool. 24.- NSP's claims that the casks can be unloaded based upon past l experience with similar casks, but this is false. NSP claims that it has based its unloading procedures on experience with similar casks, but the casks are not similar because the loading and unloading procedures are distinct. I l l

l i NSP's claims.that the TN-40 cashs can be unloaded are . baseless. 1 25. Another study of the TN-24P, conducted by INEL in 1990, also unloaded the TN-24P. Exhibit C, INEL Letter, 10/18/90; Exhibit D, INEL Letter, 11/21/90. This transfer was again an air transfer, and inapplicable for use as an example that the TN-40 can be unloaded in the pool. What study can NSP cite and produce that demonstrates that a TN-40 cask can be unloaded in a pool? Conclusions NSP has violated 10 CFR 72.122 (f) because it cannot maintain cesks. NSP has not addressed or resolved this problem and has provided inaccurate and incomplete information regarding this icsue. NSP has violated 10 CFR 72.122 (h) because the fuel is subject to d: gradation in the maintenance and unloading process specified by NSP. NSP has not addressed or resolved this problem, and has provided inaccurate and incomplete information regarding this issue. NSP has violated 10 CFR 72.122 (1) because the fuel is not retrievable, it cannot unload casks. NSP has not resolved this problem and has provided inaccurate and incomplete information rcgarding this issue. I NSP HAS violated 10 CFR $72.130 by using the TN-40 cask and fciling to make provisions that facilitate the removal of rcdioactive wastes and contaminated materials at the time the i ISFSI is permanently decommissioned. This may prevent d: commissioning in so far as a TN-40 cask that cannot be unloaded ccn therefore not be decommissioned. NSP has violated 10 CFR 572.11 by failing to provide and includ-complete and accurate material information regarding maintenance cnd unloading of TN-40 casks in their ISFSI application and in subsequent submissions regarding cask maintenance and unloading issues. NSP has received actual and constructive notice that there are cask unloading issues, has even received requests from the NRC that it address some issues, and rather than take steps to correct its unloading problem, it has instead refused to directly address these continuing problems. NSP has violated 10 CFR 572.12 by deliberately and knowingly submitting incomplete and inaccurate material information rGgarding maintenance and unloading of TN-40 casks in their ISFSI cpplication and in subsequent submissions regarding cask maintenance and unloading issues through its continual insistence that. it can unload TN-40 casks despite substantive information otherwise, and by.the knowing use of inapplicable studies to back up its false claims. NSP must be held accountable for these violations. It must not be allowed to load further casks until it has demonstrated its ability to unload them before an independent third party and has modified its Technical Specifications to reflect any changes in procedures or equipment to effect this change. Further, NSP must pay a substantial penalty for its knowing cubmission of incomplete and inadequate information regarding cask unloading issues, particularly that it is not possible to unload a cask; that no cask used for long term storage has ever been unloaded in a pool; that because necessary cask seal maintenance requires that the cask be opened, placed into the pool and submerged, which cannot be accomplished, NSP cannot properly or adequately maintain the TN-40 casks; that introducing radioactive pool water into a hot cask can cause radioactive flash steam that poses a health and safety threat to workers and the public; that introducing radioactive pool water into a hot cask can cause thermal shock that would damage cladding and assemblies and bend or warp metals with which it comes in contact; that thermal shock would impermissibly degrade

• fuel and make it irretrievable; that fuel is also irretrievable because NSP cannot unload a TN-40 cask at any time in the foreseeable future; that NSP cannot decommission the casks and site because it cannot unload the fuel to move it to another location; for these reasons, NSP has violated NRC regulations and must be substantially fined.

The NRC must prevent an erosion of public confidence in the NRC's ability to safely regulate the nuclear industry, particularly on waste management issues. The NRC must open a complete and thorough re-evaluation of dry cask storage operations ~at the ISFSI on Prairie Island and at the many other sites where the. issues raised above remain unresolved. Until such time as this ovaluation has been conducted, changes made, and problematic processes and procedures demonstrated that assure the NRC and the public of the licensee's ability to safely manage irradiated fuel in dry storage casks through the life cycle of the fuel and casks, the Materials License for ISFSI operations on Prairie Island must be suspended. During the term of suspension, no further casks shall be filled at the Prairie Island site. Dated: 4 4'I 7 ', ' g "C*<.k Exhibit A UNITED STATES .[ NUCLEAR REGULATCRY COMMISSION O wasumoron, s.c. mesma

• ...s February 25, 1997 George Crocker, Steering Committee Prairie Island Coalition F.3. Box 174 Lake Elmo, NN 55042

Dear Mr. Crocker:

As the lead manager for dry cask issues in the Office of Nuclear Reactor Regulation (NRR), Nuclear Regulatory Commission (NRC), I as responding to your letter dated January 14, 1997, to Charles Haughney. The safety analysis report (SAR) for the independent spent fuel storage installation (ISFSI) at the Prairie Island Nuclear Generating Plant provides various estimates of radiation exposure associated with the operation of the facility. Although an estimate for cask unloading is not provided, the collective dose for unloading a cask would be comparable to the estimate for loading a cask since the radiation sources and personnel activities are similar for both activities. The actual personnel exposures during the loading of seven storage casks at Prairie Island have been significantly less than the 2.315 person-rem estimate in the SAR. During discussions with the NRC staff, the licensee has stated that the personnel exposures for loading of. each of the first five casks were less than 0.27 person-rem. Regulatory limits for maximum radiation exposures to plant personnel are defined in Part 20 of Title 10 of the Code of Federal Reculations (10 CFR 20). In general, licensees are required to control the occupational dose to individual adults to less than five rems per year. The offsite release of radioactive materials during the unloading of a dry storage cask is upacted to be negligible. In regard to the worst case scenario, the SAR for the Prairie Island ISFSI includes an analysis of a hypothetical loss of confinement barrier which assumes the total inventory of radioactive gases within a cask are released. This hypothetical scenario results in a maximum individual whole body dose of 0.15 rem for a member of the public. Any credible acc.ident involving a dry storage cask at Prairie Island would result in less exposure to the general public than does this hypothetical scenario. The possible generation of steam during the refilling of a storage cask would not be a significant factor in offsite release since the steam would be vented into the spent fuel pool. In addition, the loading and unloading of casks are performed within the auxiliary building which has additional design features that minimize the release of radioactive materials. . As part of its assessments of licensees' procedures for unloading dry storage casks, the NRC staff considers the dry-run exercises performed to verify key aspects of unloading procedures, as well as licensees' actual experience in the loading and unloading of transportation casks, loading of storage casks, handling of spent fuel assemblies uiioer various conditions, and performing various activities associated with reacter facilities. En the absence of actual experience in unloading spent fuel from a cask following a long perio_d y: n, o ig

G. Crocker. of storaae. a aeneral understancino of technical canabilities and related experiences enablos the NRC staif to assess the adecuacy of a icensee's oracedures for un' cadino dry storace casks. For those examp'es of cask un' oadings mentioned in the staff's letter of January 7,1997, to Representative Jennings, the activities were performed without significant releases of radioactive material and within regulatory limits pertaining tt, occupational exposures of plant personnel. In order to ensure that the fuel assemblies in dry storage casks have maintained their integrity during storage, a gas sample is taken from the (.- early in the unloading process. In the case of Prairie Island, the licensc. unloading procedure (Enclosure 1) requires personnel to determine if additional steps or precautions are' warranted based on the analysis of the 91 sample from the cask cavity. Additional surveys and samples are taken throughout the unloading process to ensure that the radiation doses receiveo by licensee personnel are minimized. The integrity of the fuel cladding is expected to be maintained by the inert helium atmosphere during the licensed storage period of each cask. The fuel is also expected to maintain its ,,. integrity during the refilling of the cask during the unloading process. Although the limited unloading experiences with storage casks have not involved the temperature differences between fuel and coolar:t that may occur if a cask was unloaded after a period of storage, engineering evaluations and;, experiences with transportation casks have shown that " thermal shocking" is ,, unlikely to cause operational safety problems. Cask unloading would be expected to involve reflooding and opening the cask and withdrawing the fuel assemblies in a manner similar to normal fuel handling practices. In the unlikely event that fuel degradation has occurreo during storage, the unloadina may reauire additional fi tarina and even vacuumino debris from the bottom of the cask. Such steps would be dcveloped and implemented, as necessary, following the discovery of fuel damage as a result of samples and surveys required in the unloading procedure. Licensees do have experience in handling damaged fuel assemblies, including the need t< retrieve fuel pellets, as a result of several cases of fuel assembly damage that occurred during reactor operation. Although licensees would be able te develop-means to retrieve degraded fuel assemblies from a dry storage cask, th2 accumulated occupational dose to perform this activity may be increased from the previously mentioned estimates. Fuel reactivity for criticald tv conniderations could increase only under very ideal < stic and hiahly un' ikel. l' s' ntecration patterns in the fue' Unon detection that fuel disintearati2 and occurred. special measures wou' d be deve' oned and implemented to assure t adeouste safety margin is maintained durine un' ond< no. Some SARs do state that unloading is basically the reverse of loading and tt' statement, in a ger. oral sense, is true. However, such statements may tend t- 'over-simplify matters because they do not reflect that the unloading process introduces different conditions and complications compared to the loading process. In the NRC action nian for dry cask storace and related statement. made by the Ni tC staff. includina those by Mr. Kuoler. the stafd was amenasizine tsat icensees need to identify the condit ons and complicatie

G. Crocker

• that are associated with the unloadina crocess and ensure that unjoadino nrocedures address those ennemens.

The unloading procedure for tne dry storage casks at Prairie Island was inspected by the NRC staff and, following einer revisions, was found to provide adequate guidance to control the uninading process. A copy of NRC Inspection Report 50-282/95002; 50-306/95002; 72-10/95002 is provided as Enclosure 2. I trust that this information addresses your concerns. Please contact William Reckley on 301-415-1314 if you have any additional questions or concerns. Sincerely, JaHL Gail H. Marcus, Project Director Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos.: 50-282, 50-306, and 72-10

Enclosures:

As stated (2) cc w/ enc 1: The Honorable Loren G. Jennings Minnesota House of Representatives Box 27 Rush City, M 55069 cc w/o encl: see attached page 9

\\.3 UNITED STATES A. 3 NUCLEAR RE(IULATORY CCMMISSION Exhibit B wAenemeron.o.a.mesu ms \\, July 10, 1997 Mr. Roger 0. Anderson, Director Ucensing and Management lasues Northom States Power Company 414 Nicollet Mall unneapolla, Minnesota 55401

SUBJECT:

RE6UEST FOR ADDITIONAL INFORMATION ON THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2, AMENDMENT OF $ PENT FUEL POOL SPECIAL VENTILATION TECHNICAL SPECIFICATION (TAC NCS. M98782 AND M98753) Door Mr. Anderson: By letters dated May 7,1997, and supplemented May 30,1997, Northem States Power Company submitted a request to amend the Prairie Island Technical Specifications pertaining to the spent fuel pool special ventilation system. In order to review the proposed changes the staff requires some additionalinformation. Our request for additional information (RAl) is enclosed. In order to continue our review of your submittal on en expedited basis, piesse provide your response to the staff's RAI as soon as practical. If you have any / questions regardig the enntent of the RAI, piesse contset me at (301) 415-1355. Sincerely, G Both A. Wetzel, Pr Manager Project Directorate lil-l Divialon of Reactor Projects - tilAV Office of Nuclear Reactor Regulation Docket Nos. 50-282. 50 306

Enclosure:

As stated oc w/ encl: See next page q myjg z c yy

REQUEST FOR ADDITIONAL INFORMATION FOR REVIEW OF THE AMENDMENT OF THE SPENT FUEL POOL SPECIAL VENTILATION ZONE TECHNICAL SPECIFICATIONS 1. Step 8.27 of D95.2, 'TN-40 Cask Unloading Procedure" directs the cask to be filled with water. The caut8on prior to step 8.27 reeds, 'The water /steem mixture from the vont port hose may contain some radioacdve ges. The eres directly above where the hoes is discharging shall be closely monitored to determine if there is a radiological hazard." is the spent fuel pool special ventilation system operable during the performance of this step of the unloading procedure? If the spent fuel pool special ventilation system is inoperable during this step and other portions of the unloading procedurs because the overhead crane is supporting the cask through the open spent fuel pool enclosure slot doors, discuss why an inoperable ventilation system does not pose a radiological hazard and give any precautions and protections that ensure that 10 CFR Part 20 and Part 100 requirements are not exceeded. 2. Section 5.5 of the Prairie Island ISFSI [ independent spent fuel storage Installationi safety analysis report (SAR) states in part, "After moving the i cask into the fuel pool area, the cavity will be depressurised and,the cask l lowered into the spent fuel pool." However, Step 8.4 of procedure Dg5.2 directs the cask to be depressurized while it is still located in the rail bay eres. Explain the discrepancy between the two documents. Also, what is the basis for the SAR requiring the cask to be moved to the spent fuel pool area prior to depressurization? Does the SAR assume that the spent fuel pool special ventilation system will be operable during the cask l depressarisation evolution? 3. When the opent fuel cask is filled with water prior to unloading the fuel (per Step 8.27 of D95.2, 'TN-40 Cask Unionding Procedure"), discuss the likelihood that this will resutt in cracking of the spent fuel rods due to the J interaction of the cool spent fuel pool water with the hot fuel elements. If any fuel oracking is predicted, list the expected radionuotidos and quantitles that will be reisesed into the cask and into the fuel building when the oesk is vented if the filtered ventilation system is not operating during cask venting, describe how you plan to detect and prevent these radioactive gases from being released into the environment. e 1

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Exhibit E f, - f, L.* ~+ m a[ UNITED STATES 'b NUCLEAR REGULATORY COMMIS860N

• g wasuimeton, o.c. sesse.eset o....

April 16. 1997 MD4DRANDUM To: Cynthia D. Pederson, Director Division of Nuclear Materials Safety Region III L L.i 4r FROM: Jack W. Roe frecto Division o actor Projects !!!/IV Office of lear Reactor Regulation

SUBJECT:

TASK INTERFACE AGREEMENT 96-0440; DEFINING DRY CASK STOPE : TERMS (TACN05.N97346ANDM97347) 4 In response to your request dated November 26, 1996, NRR/DRPW and 9155/SFPC have discussed the questions raised and offer the following clarifications regarding the. terms ready retrieval and strudtural defects. The two basic reasons to return a cuk to the spent fuel pool and unload the spent fuel assemblies are either to (1) retrieve the fuel assemblies for further processing or disposal, or (2) respond to an event or condition that has potentf ally degraded the design reovirements established for the cask. The staff has not identifie1 the unic.adino of a cask as a renuired protective m pa.to, be samen within a soecified time in order to limit the offstte. conseque.u.cs of an accident involvino tne release of rastoachva mueriai fror. a storage cast. In regard to the reanin-~e et cask desions must allow retrieval of the spent fuel for further processina or disposal (10 un 7z.1zz( }}, the am has cansistentiy tauen a pos.itton that licensees can satisfy this requirement without maintaining the capability to retrieve tne av.. i. iusi isv. a sask within a s pesirien narico of time, ano say, ir nosessary, neveiop aisernate Jyi.ivn. ro fuel. retrieva! 1Y a cask urnoaning mmn es i- _o...iy.uvym ied idw i.o a snoruo. or snaca in a soent rusi nooi. inn is consioerec asseptacle because licensues have a great deal of flexibility in their ability to schedule and plan for the transfer of spent fuel from a storage cask to another cask for storage or shipment. ~ L Several of the actions restred by ISFSI technical specificatwas or cask certificates of compliance specify that, in the case of certate events or - conditions, a cask may need to be unloaded, or otherwise returned to a safe storage condition.- The NRC staff has stated that the Jotential need to unload a cask in response to an event or condition in the tecinical specifications or certificates of compliance does not require licensees to maintain a continuous bility to unload a cask within a specified time. This position is based on he absence of an identified event or condition involving the storage casks hat would result in an immediate threat to public health and safety. The l position is reflected in past NRC decisions such as 'the acceotability of (1) licensees not having to maintain space in spent fuel pools to accommodate CONTACT: William Reckley, NRR .l (301) 415-1314 't%2 APR t1 W ' qquaktg

(. ' C. Pederson L unloadin' g of' a cask, and (2) several licensees sharing a single cask transport vehicle between different reactor sites. In the specific case of the Prairie Island ISFSI, the NRC staff, in its safety Evaluation Report dated Jul 1993, stated that its review of the accident analyses determined that,

• se equivalent consequences, from a single cask, to any individual, from direct or indirect radiation and gaseous activity,

release after matulated accident events, dditionallyare less than tfie 50 asv (5 ree)l A in its Environmenta 1 tit establisaed in 10 CFR 71.106 b." 28,1992.th(e)staffassessedtEeaccidentdoseatthe Assessment, dated July ' site boundar{The doses are also much less than the Protective Action Guidesas found that, cstab11shed by the Environmental Protection Agency (EPA) for individuals exposed to radiation as a result of accidents. Because it has been shown that the dose equivalent to any individual from postulated accidents involving 'a single cask is below levels required for taking protective actions to protect public health, the NRC staff considers that a time-urgent unloading of a TN-40 cask is a highly unlikely event. Howevar, following certain events or conditions, the licensee is required to take corrective actions to ensure safe storage conditions and to perfom inspections to ensure a cask continues to meet applicable design requirements. This may include returning a cask to the. Auxiliary Building and/or the spent fuel pool. However, once the cask is in the spent fuel pool. it does ut have to be unloaded immediately to maint&in safe storage conditions. The licensee would have time to consider available cptions, required precautions, and other special considerations that may be involved in the required unloading of a cask. The storage methods for spent fuel mus't protect against degradation of fuel assemblies or casks that would create operational safety problems during tnloading. erational safety problems are those that involve gross rupture of the fuel e adding such that si nificant quantities of fuel material and fission products are released to he storage environments. The design requirement to maintain fuel cladding integrity during storage leads to restrictions on the fuel assemblies that can be initially loaded into the casks. Acceptance criteria for fuel' assemblies to be stored pertain to heat , generation rates, initial enrichments, assembly geometry, and other characteristics that establish boundary conditions for the analysis of fuel assembly performance during normal storage and potential off-normal conditions. The wording of prairie Island ISFSI Technical Specification 3.1.1.(6) andl the safety analysis report should be, interpreted in light of the regulatory background set forth in this paragraph. In addition, a 'TS 3.1.1.(6)- Fuel' assemblies known or suspected to have structural ' defects or gross cladding failures (other than pinhole leaks) sufficiently severe to adversely affect fuel handling and' transfer capability shall not be loaded into the cask for storage. SAR 3.1.1 ... Physical Configuration / Condition: fuel assembly shall' be intact, shall have no known cladding defects and shall not have physical damage which would inhibit insertion or removal from the cask fuel basket.

O' C. Pederson - definition for " gross cladding defect" has been incorporated into NUREG-1536, ' Standard Review Plan for Dry Cask Storage Systems," which was recently issued in final fom. In the specific case of Prairie Island, neither 10 CFR 72.122(f) or specific !$FSI technical specifications introduce additional requirements for. the fuel handling' equipment used to actually load or unload the fuel assemblies into the cask since such matte regulationsandifconses.psareaddressedunderexisting10CFRPart50 The structural requirements defined by the ISFSI SAA and technical specification are satisfied even if it is necessary to use a special handling tool to overcome problems in Iffting selected fuel assemblies, provided that these assemblies do not have gross. cladding failures and will otherwise maintain fuel assembly geometries assumed in the design- ~ basis analyses performed for the cask. The adequacy of the licensee's actions should be judged in the context of the regulat associated reactor facility operating license.fons in 10 CFR Part 50 and the If the Itcensee's actions are reasonable for the handling of fuel within the spent fuel pool, those same-actions can be credited in the determination of whether the licensee satisfies the structural integrity requirements of the ISFSI technical spot iiigation and fuel retrievability requirement of 10 CFR 72.122(1). If, on the other hand, the licensee's corrective actions are deemed inadequate.or the spect'al fuel ~handitag procedure increases the probability'of a fuel handling accident within the reactor facility, actions or inquiries from the NRC staff should be 4 presented in the context of regulations such as Appendix 8 to 10 CFR,50 or 10 CFR 50.59. The NRC Office of the General Counsel has reviewed this response and has no legal objections. Please contact William Reckley of sty staff at (301) 415-1314 if you have,any_ additional questions or concerns regarding this matter. cc(w/ incoming): C. Mehl, RI ~ B. Mallett, R!! R. Scarano, RIV a Prairie Island ISFSI Technical Specification 1.3.2, " Fuel and Cask Handling Activities," states: l Fuel and cask movement a'nd handling activities which are to be performed . in the Prairie Island Nuclear Generating Plant Auxiliary Building will be governed by the requirements of the Prairie Island Nuclear Generating Plant Facility Operating Licenses DPR-42 and DPR-60 and associated technical specifications.

[ NRC Inspecti n Report - Sierra Nuclear Ccrp. i l j Selected Reoorts Index l News and Information j NRC Home Pane i E-mmi 1 l April 15,1997 Mr. Art J. McSheny President Sierra Nuclear Corporation One Victor Square Scorts Valley, CA 95066

SUBJECT:

NRC INSPECTION REPORT NO. 72-1007/97-204 AND NOTICE OF NONCONFORMANCE l

Dear Mr. McSherry:

i This letter refers to the inspection conducted March 17-21,1997, at your facility in Scotts Valley, California, and tt two ofyour fabrication contractors' facilities: March Metalfab, Inc., in Hayward, California; and N r-Cal Metal Fabricators, in Oakland, California. The team examined information about. seal weld failures on dry spent fuel storage casks at the Palisades and Arkansas Nuclear One (ANO) nuclear power plants., Additi:nally, the team assessed the adequacy of your corrective actions taken for the findings identified in Inspection Reports 72-1007/M-204 and %-208, regarding the Model VSC-24 dry spent fuel storage system manufactured under Certificate of Compliance No. 72-1007. The enclosed report (Enclosure 1) presents the results of our inspection. The team held an exit meeting with you in the Sierra Nuclear Corporation offices on March 21,1997. During the inspection, the team found that you failed to meet certain Nuclear Regulatory Commission requirements. The team identified four nonconformances regarding failures to perform work in accordance with your Quality Assurance Program. The nonconfonnances were failures to (1) examine the potential generic aspects of the shield lid weld failures at ANO and Palisades, (2) submit a change to the Certificate of Compliance to correct a nonconservative requirement for the drain-down time limit for a loaded cask, (3) submit a Safety Analysis Report change to correct the 1986 American Society of Mechanical Engineers Code omission of nondestmetive examination requirements for temporary wehments, and (4) control measunng Mest equipment. 1 Two of the nonconformances raise safety concerns. First, the shield-lid weld failures affect the integrity of a cask confinement boundary. The root-cause of the shield-lid failures and the potential for delayed cracking on loaded casks must be understood. Although the failure of both the cask's 'mner shield-lid seal weld and outer structural-lid weld would not pose an off-site threat to public health and safety, such an occurrence would cause the loss of the helium atmosphere inside the cask. This loss could resuh in cladding degradation and future fuel handling and retrievability problems. Since one of the design requirements of the cask is the long-term protection of the fuel cladding [10 CFR 122(h)], such degradation would be unacceptable. Second, the nonconservative Technical Specification for cask drain-down time affects the margin to criticality. N Gjo i .ic'.i n M

Page 2 Sierra Nuclear Corporation's lack cf timely and comprehensive action, in dealing with these imponant safety issues, is a dgni6 cant regulatory concern. As the ceni6cate holder, Sierra Nuclear Ccrporation is responsible for the adequacy of the design ofits fuel storage casks. We expect Sierra Nuclear to take a central role in resolving each technical problem associated with your cask design. We have arranged a meeting with you on May 6,1997, to discuss this matter further. This meeting is open for public observation. At the meetmg you should be prepared to discuss your shon term and longer term corrective actions to address the issues and concerns raised by our layh. Please provide us, within 30 days from the date of this letter, a written statement in accordance with the instructions specified in the attached Notice of Nonconformance (Enclosure 2). We will consider extending the response time ifyou can show good cause for us to do so. In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its enclosures, and your response will be placed in the NRC Public Document Room (PDR). Sincerely, . /dgned/ Susan Frant Shankman, Chief Transportation Safety and Inspection Branch Spent Fuel Project Office, NMSS

Enclosures:

t

1. Inspection Repon 72-1007/97-204

2. Notice of Nonconformance Docket No. 72-1007 4

i

]:Co5firmat:ry Actinn Letter-Arkansns Nuclear ~ Selected Reports index l News and Infannation l NRC Home Page l E-mail May 16,1997 CAL No. 97 7-002 Mr. C. Randy Hutchinson Vice President, Operations ANO Entergy Operations,Inc. 1448 S. R. 333 AR 72801

SUBJECT:

CONFIRMATORY ACTION LETTER

Dear Mr. Hutchinson:

During the week of March 17,1997, U.S. Nuclear Regulatory Commission staffinspected Sierra Nuclear Ccrporation (SNC) and two ofits fabrication contractor facilities. SNC holds Certificate of Compliance No. 1007 for the VSC-24 dry storage cask. This inspection focused on welding problems with VSC-24 casks. used at the Palisades and Arkansas Nuclear One (ANO) nuclear power plants. The problems were in the welds joining the cask shield lid to the multi-asaembly storage basket (MSB). The Pali welding dif5culty occurred in March 1995 and the first instance at ANO in December 1996. AAer the SNC inspection, problems arose while welding another ANO cask on March 26,1997. NRC is concerned about the difficulties , e.,r!!;p.ggn,tered with the welds joining the shield lid to the MSB, since this weld is part of the confinement boundary of'th'e V$C fGurtherm6re~,'the weld between the MSB and the structurallid 'may be susceptible to the same faifure miichiihisms,3._thubigid_Ed weld. It is possible that these cartici lar weld problemq may s not develon until after c=<le welds have underaone non-destructive exarhihavian. Alt mugh'such weld :adures would not pose an off-site threat to public health and safety, such an occurrence would cause the loss of the ~ ~ ' helium atmosphere inside the MSB. This cpodition could_r_esukiAnc}&degradatiWarii fiinTre fuel [Iandling and retrievabdity problems. The March 1997 inspection revealed that neither SNC nor the user licensees had performed a comprehensive root-cause analysis of the first two weld problems. An understanding of the root cause is essential to preventing recurrence when welding future casks, and to assessing the possibility of additional weld Prh. perhaps nadaartad or delayed, in loaded casks. On May 6,1997. NRC held a public meeting with SNC representanves to discuss SNC's implemented and planned actions in response to the weld problems and inspection Wiar Representatives of your staff attended this meeting. As stated at this meeting, the staff remains concerned that the root cause(s) of the weld problems have not been conclusively determmed. Pursuant to a May 14,1997, telephone conversation between Randy Edington and Charles Haughney, Deputy Director of the Spent Fuel Project OfEce, OfEce of Nuclear Material Safety and Safeguards, it is our understanding that you will take the following actions before loading additional VSC-24 casks with spent i nuclear fuel:

UNITED STAYa5 3 NUCLEAR REGUULTORY COMMLSSION .g nEGCN O - amtwannsm usnono USLE, a1JMole S1834351 June 30, 1995

• • o**

Mr. E. Watz1, Vice President Nuclear Generation Northern States Power Company 414 Nicollet Mall Minneapolis, MI 55401

Dear Mr. Watz1:

This refers to the special NRC inspection from January 24 through May 11, 1995, of dry cask storage activities at the Prairie Island site. This inspection was conducted by the resident inspectors, selected RIII based inspectors, and technical staff from the Office of Nuclear Reactor Regulation and the office of Nuclear Materials saf'ety and safeguards. The purpose of this inspection was to evaluate the acceptability of the as-built TN-40 cask and to assess your performance relative to dry cask storage including the arcocerational testina activities. We discussed the results of this inspection with you and other members of your staff at a public exit meeting on April 28, 1995. At that meeting we identified five' items that required further resolution. You provided us with additional information for~each of these items and we completed our review of the subject items during the next two' weeks. On May 11, the NRC issued a schedular exemption from the requirements of 10 CFR Part 72.82(e) allowing you to submit the results of your preoperational test less than 30 days before the receipt of fuel at your onsite Independent Spent Fuel Storage Installation. On May 12 you loaded the first cask with spent fuel. The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel. Based on the results of this inspection, we concluded that you were ready to , safoly load spent' fuel into the TM-40 dry storage cask and transport this cask to the onsite 15FSI. We also did not identify any safety concerns'with the subject cask. However, one violation of NRC requirements was identified during the course of this inspection, as specified in the enclosed Notice Inf Violation (Notice). This violation pertained to cask handling, loading, and unicading activities that were not prescribed by procedures of a type appropriate to the circumstances. Although 10 CFR 2.201 requires you to submit to this office, within 20 days of yrur receipt of this Notice, a written statement of explanation, we note that this violation had been corrected and those actions were reviewed during this inspection. Therefore, no response with respect to this violation is required. However, we are disappointed that NRC inspectors, rather than your own staff, identified these procedural deficiencies. l Ink O y j)

Isrst saa e TABLE 5.1-2 (Continued) ANTICIFATED TIME AND FER50lOfEL RIQUIRIMENTS FOR CASK HANDLING OPERATIONS No. of Time Avg. Distance ggyration Personnel M (ft) from Cask I { Starmaa Aram .23. Unioad from vehicle position in location 5 60 5 (D1, D2, D3) 24. Check surface dose race (D6) 5 30 3 25. Connect pressure instrumentation (D4, DS) 5 30 5 Feriodie Maintan=nre 1. Visual surveillance (NA) 2 15 5 2. Repair surface defects (NA) 2 60 3 3. _ Instrument testing and calibration (NA) 2 180 5 6. Instrument repair (NA) 2 60 3

1. Maior Maintenanes y

(cnce in 20 years) 3 1950** 8 1. Replace cask lid seals s Therefore, the number Eo measurable dose associated with this activity. 0 of personnel, time and distance are not significant. Parenthetical information corresponds to Table 5.1-1 activity numbers. O Total time to transfer cask to spent fuel pool, replace lid seals, and l return cask to ISTSI pad. REV. 2 9/91 TABLE 5.1 2

NOTICE OF VIOLATION Northern' States Power Company Dockets No. 50-282; 50-306; 72-10 Prairie Island Nuclear Plant Licenses No. DPR-42; DPR-60; SNN-2506 During an NRC inspection conducted from January 24 through May 11, 1995, a violation of NRC requirements was identified. In accordance with the " General Statement of Policy and Procedures for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, the violation is listed below: 10 CFR Part 72.142(b) requires a licenses to establish, maintain, and execute e' a quality assurance (0A) program with regard'to an Independent Spent Fuel i Storage Installation (ISFSI Subpart 6, " Quality Assuranc)e."that satisfies each of the applicable criteria of In sieeting the Part 72.142(b) requirement, 10 CFR Part 72.142(d) accepts a Commission-approved quality assurance program which satisfies the applicable criteria of Appendix 8 to 10 CFR Part 50. As srch, the ISFSI Safety Analysis Report ' states that the previous 1r naaroved Northern States Power QA program wsich satisfie's applicanie criteria of 10 CFR Part 50, Appendix 8,'will ne applied to activities, structures, systems, and components of the ISFsI connensurate witn sneir importance to safety. Criterion V of Appendix 8 to 10 CFR Part 50 requires that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and that these activities be" accomplished in accordance with the associated instructions, procedures, or, drawings. Cask handling, loading, and unloading are activities affecting quality. 9tttrary to the above, cask handlina. Inadina. and unlandinn activities were Sotyra<mnan nv annrovea crocedur** af 2 tvna aoorooriate to the ~ muustancesjas evidenced by therrollowing examples:j 1. ,$urveillance Procedure. SP 1077, "Special Lift Fixture for the TN-40 cast," d' d not address dimensional checks of the special lifting device, as required. 2. Surveillance Procedure, SP 1075, "TN-40 Fuel Selection and Identification," did not incor> orate the requirement of Technical Specification (TS) 4.1.2, whics states that "before inserting a spent fuel assembly into a cask.... the identity of each fuel assembly shall be independently verified and oocumented." 3. Procedure D95.1, 'TN-40 Cask Loading Procedure pecified in the prerequisites section that SP 1077 be perform ays prior to loading a cask. However, the TS 4.'19-requirement to pe ora a visual i==a -tion of the lifting device (lift beam and extension) aN v=rify operability] of the device 7 days prior to use, was not identif' eu in un.a. muru also was no procedure identifying actions required to verify operability of the lifting device. I Q4=64)M) & I

While the inspectors recognized that finalizing the loading and o unloading procedures was contingent upon completion of the dry run and the subsequent incorporation of any lessons learned, there were many aspects of the procedures which should have been in place oerore tne dry ror exampie, lechnicas apearication requirements were nos ectively incorporated into the loading and unloading procedures I (paragraph 3.2). In addition, the licennee dk not como ete rev' ew and e _ approval of the unloading proceoure unti: the c ay follow' ng subs: ssfon / or rne oreoperau onas test report. Subnission of t ut renart mo t Ted that the licensee was raarfy to loaa a cast with spent fuel and / subsequently un oad the cask, if necessary. ~ The licensee did not take a disciplined approach to inspecting the fuel o designated for cask storage as evidenced by weaknesses identified by the inspectors during observation of fuel inspection activities (paragr,aph 7.3). Some weaknesses were noted with 'the licensee's documented basis for o safety evaluation conclusions (paragraph 8.2). e f 3 i

s." Cohfirmctory Acti:n Letter-Araansas Nuclear .[*s ,., man er Suected Reports Index l News and Information l NRC Home Page l E-aml ' May 16,1997 CAL No. 97-7-002 Mr. C. Randy HutcMaana Vice President, Operations ANO Entergy Operations,Inc. 1448 S. R. 333 RussellviDe, AR72801

SUBJECT:

CONFIRMATORY ACTION LETTER

Dear Mr. Hutchinson:

During the week ofMarch 17,1997, U.S. Nuclear Regulatory Commission staffinspected Sierra Nuclear Cerporation (SNC) and two ofits fabrication contractor facilities. SNC holds Certificate of Compliance No. 1007 for the VSC-24 dry storage cask. This inspection focused on welding problems with VSC-24 casks. used at the Palisades and Arkansas Nuclear One (ANO) nuclear power plants. The problems were in the w;lds joining the cask shield lid to the multi-assembly storage basket (MSB). The Palisados welding dif5culty occurred in March 1995 and the first instance at ANO in December 1996. After the recent SNC inspection, problems arose while welding another ANO cask on March 26,1997. NRC is concerned about the difficulties e_!! San,te, red with the weldsjoining the shield lid to the MSB, since this weld is part of the confinement boundary ofilie V$CW Yurthermore','the welif between the MSB and the s'tructurailid 'may be susceptible 13 the same falfure_miichEnismsy.thg_3hije { lid weld. It is DOssible that these cartb lar weld oroblem1 may n:t develon until after cask welds have underaone non-destructive =msHatran7 Jt mugh'such weld :adures woukt not pose an off-site threat to pubhc health and safety, such an occurrence would cause the loss of the helium a_tmosphere inside the MSB. This conditiOLgoulir_esukin fu_elcjadg@Warid_ftitMel liandimg and retrievability problems. q i The March 1997 inspection revealed that neither SNC nor the user licensees had performed a comprehensive root-cause analysis of the first two weld problems. An understanding of the root cause is essential to preventing recurrence when Minis future casks, and to assessing the possibility of additional weld prahla==. perhaps unhactad or delayed, in loaded casks. On May 6,1997 NRC held a public meeting with SNC i r,-zr*ves to discuss SNC's implemented and planned actions in response to the weld problems and j inspection findings. Representatives ofyour staff attended this meeting. As stated at this meeting, the staff j remains concerned that the root cause(s) of the weld problems have not been conclusively determined Pursuant to a May 14,1997, telephone conversation between Randy Edington and Charles Haughney, Deputy Director of the Spent Fuel Project OfEce, OfEce of Nuclear Material Safety and Safeguards, it is our understanding that you will take the following actions before loading additional VSC-24 casks with spent nuclear fuel: i 77;gm;Typr}/4-(/gg

~ E6nfirmatory Acti:n Letter-Arkanses Nuclear Selected Reports Index

j News and InformatiSD j NRC Home Pane lW May 16,1997 CAL No. 97-7-002 Mr. C. Randy Hutchinson Vice President, Operations ANO Entergy Operationa,Inc.

1448 S. R. 333 - 3 Russellville, AR 72801

SUBJECT:

CONFIRMATORY ACTION LETTER

Dear Mr. Hutchinson:

During the week ofMarch 17,1997, U.S. Nuclear Regulatory Commission staffinspected Sierra Nuclear - Corporation (SNC) and two ofits fabrication contractor facilities. SNC holds Certi6cate of Compliance No. 1007 for the VSC-24 dry storage cask. This inspection focused on welding problems with VSC-24 casks used at the Palisades and Arkansas Nuclear One (ANO) nuclear power plants. The problems were in the ' w: Ids joining the cask shield lid to the multi-assembly storage basket (MSB). The Palisades welding dif5culty occurred in March 1995 and the nrst instance at ANO in December 1996. AAer the recent SNC inspection, problems arose while welding another ANO cask on March 26,1997. NRC is concerned about the difficulties en._c_o.un_tered with the welds joining the shield lid to the MSB, since this weld is part of the confinement ~ boundary of the VSC'24'. Furthermore,~the weld between the MSB and the structural lid may be susceptible ta the same faifure michisisms_as_thubiglilid weld. It is possible that these partiet lar weld problema may ~ nit develon un~til afteTcask welds have underaone non-c estructive exarhisavian. Alt sough'such weld :adures ~ wruld not pose an off-site threat to pubhc health and safety, such an occurrence would cause the loss of the - helium atmosphere inside the MSB. T,his conditign_could_rgsuit,i3n[_cWging desdftioTan'd,gtGe[ej ,Tandling and retrievability problems. The March 1997 inspection revealed that neither SNC nor the user licensees had performed a comprehensive root-cause analysis of the Erst two weld problems. An understanding of the root cause is essential to preventing recurrence when welding future casks, and to assessing the possibility of additional weld problems, perhaps undetected or delayed, in loaded casks. On May 6,1997, NRC held a public meeting with SNC representatives to discuss SNC's implemented and planned actions in response to the weld problems and EY+2 ion Sndings. Representatives ofyour staff attended this meeting. As stated at this meeting, the staff remains concerned that the root cause(s) of the weld problems have not been conclusively determined Pursuant to a May 14,1997, telephone conversation between Randy Edington and Charles Haughney, Deputy Director of the Spent Fuel Project OfEce, Of5ce of Nuclear Material Safety and Safeguards, it is our understanding that you will take the following actions before loading additional VSC-24 casks with spent ~ nuclear fuel:

- (1) Detennine that your weldirig and inspection practices p' ovide reasonable assu: ance that crackmg, including possible undetected or delaved crackms, cell out occur in the welds m!ing the shield lid and 2 structurallid tothe M5B If necessary modify yourweldirig p pcesses to inhint recurrence cf these welding Problems (2) On completon of this activu, and an.lems.td raays befuie lundirg a:ene VSC-24 cask with spent fuel, you will submit to the Directot, Offwe of Nucmat.Wrate6al Safety and tidegods, a written description of any procedural or design modakatimenade #nh reysr3 w.bm.'. The sutmittal should include the technicaljustification Er eacb nudifaatn.m 4 opj cd she wmi rail snould % sent to William F. Kane, Director, Spent Fuel Propa 05ce, ard to iam Tegime Administra :r (En may include in this response the infonnation required by inero2 bemw, mt.fmes ar Je aciloo requireCoy item I above.) Pursuant to Section 182 of the Asouuc Energy Act,42 U iC 2232,;m we required to: (1) Notify me immediately if your understanding differs from that set fonh above; (2) Notify me in wnting when you have w,.Aed the actions addressed in this Confirmatory Action Letter. Issuance of this Confirmatory Action Letter does not preclud issuance of an order formalizing the above t commitments or requiring other licensee actions; nor does it preclude NRC firom takmg enforcement action f r vi:lations of NRC requirements that may have prompted the issuance of this letter. In addition, failure to take the actions addressed in this Confirmatory Action Letter may result in enforcement action. In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter,and your response will be placed in the NRC Public Damment Room (PDR). To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR - without redaction. However, if you find it necessary to include such information, you should clearly indicate the speci6c infonnation that youdesire not to be placed in the PDR, and provide the legal basis to support your request for withholding the information from the public. Sincerely,- Malcolm R. Knapp, Deputy Director Office ofNuclear Material Safety and Safeguards Dockets 72-1007, 72-13, 50-313, 50-368

1 .u.... o s* a, ISFSI SAR TABLE 5.1-2 ANTICIPATED TIME AND FIRS 000tEL REQUIRDIDITS FOR CASK HANDLING OPERATIONS Onoration No. of Time Avg. Distance Parsonnel fainl (ft) from cask Receivina 1. Unioeding (A1) l 2. Inspection (A2 through A7) j 3. Transfer to cask loading pool (AS) t'= = k ? = = 4i n e Pool J 4. Iower cask into pool (51) 5. Lead fuel (32 through 54) 5 i 6. Flace lid on cask (35) 5 7. 1.ift cask to pool surface (56) 5 30 5 8. Install lid bolts (36) 5 120 3 9. Drain cavity (37 through Bil) 5 90 6 10. Transfers to decontamination area (512) 3 60 10 Decone==ination Area V 11. Decontaminate cask (C1, C2) 3 120 3 12. Remove vent plugs 2 30 5 13. Drying, evacuating, backfilling (C3 through C13) 2 480 5 1$. Install top neutron shield C14) 2 15 3 15. Install pressure transducers (C15 through C17) 2 30 5 16. Pressurize interspace (C18) 17. Check Isakage (C19) 2 30 5 18 Check surface temperature (C20) 2 30 5 19. Check surface dose rate (C21) 2 30 3 20 Install protective cover (C22) 2 30 5 21. Imad on transport vehicle (C23) 3 60 5

22. Transfer to storage area C24) 3 60 10 i

l TABI2 5.1-2 REV. 2 9/91 I

\\ 1 e ISFSI SAR TABLE 5.1-2 (Continued) ANTICIPATED TIME AND FERSONNEL REQUIRDENTS FOR CASK HANDLING OPERATIONS Onsration No. of Time Avg. Distance Personnel M (ft) from Cask ( l Storare Area 23. Unioad from vehicle position in location (D1, D2, D3) 5 60 5 24. Check surface dose race (D6) 5 30 3 25. Connect pressure instrsamentation (D4, D5) 5 30 5 Feriodic Maintenance 1. Visual surveillance (NA) 2 15 5 2. Repair surface defects (NA) 2 60 3 3. Instrument testing and calibration (NA) 2 180 5 4 Instrument repair (NA) 2 60 3 / s t 3 1950** 8 o No measurable dose associated with this activity. Therefore, the z. umber of personnel, time and distance are not significant. O Parenthetical information corresponds to Table 5.1-1 activity numbers. Total time to transfer cask e replace lid seals, and 8 return cask to ISTSI pad. TABLE 5.1-2 REV. 2 9/91

a '. * ' '. ? ' ISFSI SAR TABLE 5.1-2 (Continued) ANTICIFATED TIME AND FIRSomtEL REQUIRDGNTS FOR CASK MANDLING OPERATIONS Oneration No. of Time Avg. Distance Farsonnel M fft) from t'==k Storama Area

23. Unload free vehicle position in location (D1, D2, D3) 5 60 5

24 Check surface dose rate (D6) .5 30 3 25. Connect pressure instrumentation (D4, DS) 5 30 5 Periodic Maintenance 1. Visual surveillance (NA) 2 15 5 2. Repair surface defects (NA) 2 60 3 3. Instrument testing and calibration (NA) 2 180 5 4. Inscrumsat repair (NA) 2 60 3 /Maior Maincanance j. (ence in 20 years) 1. Replace cask lid seals 3 1950** 8 O No measurable dose associated with this activity. Therefore, the number of personnel, time and distance are not significant. o Parenthetical information corresponds to Table'5.1 1 activity numbers. Total time to transfer cask to spent fuel pool, replace lid seals, and return cask to 18FSI pad. TABLE 5.1-2 REV. 2 9/91

cperattoral checks of v:hicle brakes, lifting soutpment, turntatdes,, jacks, i and cask links. 3.1.5 Surveillance Needure. SP 1075. 'TN-40 Fuel Selection and i Jdentification' r - The inspectors reviewed SP 1075 and the cask loading procedura, D95.1, to verify that selected Technical Specification (TS) requirements had been incorporated into procedures. Surveillance requirements for ensuring that fuel assemblies which satisfy the criteria of TS 3.1.1 would be loaded into .the cask, are defined in TS 4.1.- required that, " fuel assemblies known or suspected to have TS 3.1.1(6) defects or gross cladding failures (other than pinhole leaks)ility structural sufficiently severe to adversely affect fuel handling and transfer capab The licensee originally - shall not be loaded into the cask for storage.' int:nded te visually inspect fuel assemblies designated for loading with bicoculars to identify any ' structural defects or gross cladding failures." The. inspectors questioned the efficacy of this' technique to provide a thorough inspection of the fuel. After further discussion with Region !!! staff on -fuel inspection techniques, the licensee elected to use video recording equipment to perform the fuel inspection. The inspectors considered tiis a preferable method for identifying fuel anomalies and ensuring compliance'with The inspectors observed portions of the actual fuel inspection and TS 3.1.1. identified weaknesses with the licensee's approach to this activity as discussed in paragraph 7.3. During the review of SP 1075, the inspectors identified that the procedure did tot incorporate the requirement of TS 4.1.2, which stated that "before inserting a spent fuel assembly into a cask... the identity of each fuel The inspectors assembly shall be independently verified and documented." discussed the independent verification requirements of TS 4.1.2 with the licensee. Subsequently, the licensee revised SP 1075 to address independent verification of fuel assembly identification. Based on observations of the actual' fuel inspection, the inspectors. concluded that the licensee met all TS The failure to incorporate the requi ements for fuel identification. requirements of TS 4.1.2 into SP 1075 is considered an example of a violation of Criterion V of Appendix B to 10 CFR Part 50 (50-282/95002-01; 50-306/95002-01; 72-10/95002-01(DRP)). 3.2 Leadina and unloadina Procedures The ins >ectors reviewed the loading (D95.1) and unloading (095.2) procedures fcr tecinical adequacy and to determine if the lessons learned from the preoperational testing / dry run had been appropriately incorporated into the . procedures.- 3.2.1 D95.1. 'TN-40 Cask loadina Procedure" Th3 original D95.1 procedure specified in the prerequisites section that SP -1077 be perfomed 30 days prior to loading a cask. However, the Technical Sp:cification (TS) 4.19 requirement to perfom a visual inspection of the 10 i i

a. t [7590-01-P] NUCLEAR REGULATORY COMMISSION- - 10 CFR Part 72 (Docket No. PRM-72-4) g. Prairie Island Coalition;- Receipt of Petition for Rulemaking AGENCY: Nuclear Regulatory Commission. - ACTION: Petition for rulemaking; Notice of receipt. I

SUMMARY

The Nuclear Regulatory Commission (NRC) has received and requests public comment on a petition for rulemaking filed by the Prairie Island Coalition. The -

petition has been docketed by the Commission and has been assigned Docket No.- PRM-72-4. The petitioner requests that NRC undertake rulemaking to examine certain . issues addressed in the ' petition relating to the potential for thermal shock and corrosion 'in dry cask storage.' The petitioner requests that the NRC amend its regulations that . govem independent storage of spent nuclear fuel in dry storage casks to define the parameters of acceptable degradation of spent fuel in dry cask storage. The petitioner - also requests an amendment to the regulations to define the parameters of retrievability of spent nuclear fuel in dry cask storage and to require licensees to demonstrate safe cask unloading ability before a cask may be used at an Independent Spent Fuel 1 Storage installation (ISFSI). I i $WOcLg5 %,o f

~* .. ~ 2 . DATE: Submit comments by (75 days following publication in the Federal Register). - Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except as to comments received on or before this date.- ) ADDRESSES: Submit comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Rulemakings and Adjudications staff. l ) Deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:30 am and 4:15 pm on Federal workdays. For a copy of the petition, write: David L. Meyer, Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. You may also provide comments via the NRC's interactive rulemaking website through the NRC home page (httpi//www.nrc. gov). This site provides the availability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Ms. Carol Gallagher, 1 (301) 415-5905 (e-mail: CAG@nrc. gov). FOR FURTHER INFORMATION CONTACT: David L. Meyer, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone: 301-415- + 7163 or Toll Free: 1-800-368-5642 or E-mail: DLM1@NRC. GOV. t ..,4

'..1 ' 9' 3 SUPPLEMENTARY INFORMATION:

Background

The Nuclear Regulatory Commission received a petition for rulemaking submitted by George Crocker on behalf of the Prairie island Coalition (PIC) in the form of a letter and an attached document addressed to L. Joseph Callan, Executive Director for Operations, NRC, dated August 26,1997. Most of the issues presented in Mr. Crocker's letter and the attached ' document pertain to a petition filed under 10 CFR 2.206 regarding dry storage cask regulations that has been reviewed by the NRC Office of Nuclear Reactor Regulation (NRR). See 62 FR 53031. The resolution of these issues is presented in a decision published by the Director, NRR (DD-98-02; 2/11/98). This notice pertains to paragraphs 13,14, and 15 on page 3 of the document attached - to the August 26,1997, letter from PIC. These paragraphs contani a request for I rulemaking under 5 U.S.C. 553(e) of the Administrative Procedure Act (APA). The NRC has determined that the issues presented in paragraphs 13,14, and -.15 of the PIC document constitute a petition for rulemaking under 10 CFR 2.802. Paragrsph 13 requests NRC to solicit and review information regarding thermal shock an'd corrosion inherent in dry cask storage and usage and to define the parameters of degradation of spent nuclear fuel in dry cask storage acceptable under 10 CFR 72.122(h). Paragraph 14 requests NRC to define the parameters of retrievability required under 10 CFR 72.122(l). Paragraph 15 requests NRC to require demonstration of a safe cask unloading ability before a cask may be used at an ISFSI. These requests do meet the sufficiency requirements for a petition for rulemaking under . v. f

4 10 CFR 2.802. The petition, consisting of paragraphs 13,14, and 15, has been docketed as PRM-72-4.' As set forth in the petition, the petitioner is the Prairie Island Coalition (PIC), a consortium of environmental, business, citizen, and religious groups, and tribal and urban Indian organizations. PIC is involved in locating and disseminating information ] regarding dry cask storage of spent nuclear fuel, and opposes Northem States Power Company's (NSP) plans to construct and operate an ISFSI at the Prairie Island Nuclear Generating Station (PI). PIC has participated in various Minnesota and NRC proceedings that pertain to operational and waste issues at the Prairie Island facility. The NRC is soliciting public comment on the petition for rulemaking submitted by the Prairie Island Coalition that requests the changes to the regulations in 10 CFR Part 72 discussed below. Discussion of the Petition The petitioner notes that the regulations in 10 CFR Part 72 establish requirements and criteria for spent fuel dry cask storage and usage. The petitioner has requested a rulemaking proceeding to examine issues regarding degradation, retrieval, . and unloading of spent nuclear fuel in dry storage casks. Dearadation of Soent Nuclear Fuel .The petitioner requests an amendment of the regulations in 10 CFR Part 72 to j i define the parameters of spent fuel degradation that are acceptable to the NRC under 10 CFR 72.122(h). Section 72.122(h) provides that spent fuel cladding must be protected during storage against degradation or that the fuel must be configured such

5 that degradation will not pose an operational safety concem. The petitioner is concemed about the potential effect of spent fuel degradation on the ability to safely unload a dry storage cask. The petitioner believes that factors such as thermal shock will cause spent fuel to degrade in the course of unloading and expose onsite personnel and the environment to radioactive emissions. The petitioner states that no procedures have been developed to protect operational safety and to assess worker or offsite radiation exposure in such a situation. The petitioner cites a February 25,1997, letter from Dr. Gail H. Marcus, NRC, to Plc in support of the petition. PIC asserts, based on the letter, that temperature differences between spent fuel and coolant create the potential for thermal shock and spent fuel degradation. PIC also believes the TN-40 cask is subject to failed welds and to fuel 4 degradation due to cask seal failure as a result of helium gas release. PIC cites as i support for the petition a letter dated April 15,1997, from Dr. Susan Frant Shankman, NRC, to Sierra Nuclear, and contends that cladding degradation during storage is unacceptable because it could lead to future fuel handling and retrievability problems. The petitioner also cites the Safety Analysis Report submitted by NSP for the ISFSI at i . the PI facility that requires the liconsee to replace cask seals to prevent a helium leak and fuel degradation. Copies of the supporting documents referenced above are attached to the petition. i l Plc contends that NRC has not adequately addressed the possibility of damage caused by thermal shock when cool water from a storage pool is placed in a cask that contains spent nuclear fuel. The petitioner also contends that NRC had not adequately

~ 6 . addressed degradation of spent nuclear fuel due to the loss of helium from failed seals or due to the passage of time. Retrievability of Soent Nuclear Fuel The petitioner also requests an amendment to the regulations in 10 CFR Part 72 that govem storage of spent nuclear fuel in dry storage casks to define the parameters of retrievability of spent fuel required by the NRC under 10 CFR 72.122(l). Section 72.122(l) provides that spent fuel storage systems must be designed to allow ready retrievability of the spent fuel for future processing or disposal. PIC is concemed that the NRC has not taken into account the potential problems that may be encountered in unloading a cask to retrieve spent fuel. In support of its claim, PIC. cites an April 16,1997, memorandum from Jack Roe, NRC, to Cynthia Pederson, NRC Region Ill, and asserts that this memorandum is evidence that - NRC has not taken into account possible problems with retrieval of spent fuel. The petitioner also cites a study of the TN-24 cask conducted by the Idaho 'Natic nal Engineering Laboratory (INEL) in 1990, which involved opening TN-24 casks that contained canisters of spent fuel assemblies that had been stored for several years. The petiboner contends that the INEL study found the thermal damage so great that some canisters containing spent nuclear fuel could not be retrieved from the cask. The pehtioner believes that the INEL study and the cited NRC memorandum, copies of which are attached to the petition, demonstrate that spent nuclear fuel cannot be reliably retrieved from dry storage casks. I

p t s. 7 Unloadina of Soent Nuclear Fuel Lastly, the petitioner requests an amendment to the regulations to require licensees to demonstrate the ability to unload spent nuclear fuel safely from a dry storage cask before a cask may be used at an ISFSI. The petitioner contends that if a ) licensee can demonstrate ability to unload spent nuclear fuel safely from a cask in a pool after long-term storage, then the public will have assurance that a spent fuel storage cask can be unloaded. L ' PIC contends that a cask may need to be unloaded for various reasons.. The petitioner notes that Minnesota law in, in the Matter of Soent Fuel Storage Installation. 501 N.W.2d 638 (Minn. Ct. App.1993), requires a licensee to move casks after eight years of temporary storage. The petitioner believes that the 1990 NRC Waste [ Confidence Decision also contemplates that casks will need to be unloaded before l transport to a Federal interim site or repository. Plc believes that although NRC regulations do not require a licensee to be able to immediately unload a cask, NRC clearly requires a licensee to be able to unload the j spent fuel at some point. The petitioner also believes that because in-pool unloading of 1 apont fuel from a dry storage cask that has contained the fuel for a protracted time period has not been completed, there is sufficient reason to require a licensee to demonstrate the ability to actually unload a dry storage cask underwater. PIC states that it would be satisfied if a licensee can demonstrate the ability to unload spent j nuclear fuel from a dry storage cask at some reasonable point in time. i (.-'

i r* l 8 The Petitioner's Conclusions l l The petitioner has concluded that NRC regulations in 10 CFR Part 72 that govem independent storage of spent nuclear fuel in dry storage casks must be amended. Plc has concluded that thermal shock and associated degradation of spent nuclear fuel during the unloading of dry storage casks has not been adequately addressed in NRC regulations. The petitioner requests an amendment to the l regulations to define the parameters of acceptable degradation of spent nuclear fuel in dry storage under 10 CFR 72.122(h). The petitioner has also concluded that NRC regulations do not adequately. address issues related to the retrieval of spent nuclear fuel from dry storage casks. The l petitioner requests an amendment to the regulations to define the parameters of retrievability of spent fuel from dry storage casks required under 10 CFR 72.122(l). i Lastly, the petitioner has concluded that NRC regulations do not adequately address issues pertaining to unloading of spent nuclear fuel from dry storage casks. - The petitioner requests an amendment to the regulations to require licensees to demonstrate the ability to unload spent nuclear fuel safely from a dry storage cask before the cask may be used at an ISFSI. Dated at Rockville, Maryland, this day of March,1998.- l For the Nuclear Regulatory Commission. John C. Ho Secretary of the Commission.

s.

• , o CONGRESSIONAL CORRESPONDENG SYSTEM DOCUMENT PREPARATION CBECEIST This check list is to be submitted with each docxement (or gmup of Qs/As) sentfor pmcessing into the CG.

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1. BRIEFDESCRIPTION OFDOCUMENT(5)

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2. TYPE OFDOCUMENT X CORRESPONDENG HEARINGS (Qs/As)

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