ML20216G282
| ML20216G282 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/26/1997 |
| From: | Crocker G AFFILIATION NOT ASSIGNED |
| To: | Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| References | |
| FRN-62FR12040, FRN-63FR12040, RULE-PRM-72-4 PRM-72-4, NUDOCS 9803190320 | |
| Download: ML20216G282 (41) | |
Text
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P.O. Box 174 - take Elmo, MN 55042 - Phone: 612 770 3861 FAX 770 3976 g- . ...
August 26,1997 ,
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DOCKET NUMBER cecerra M PETITION RULE PRM 74d m 1d W [T (wwao) : menomsm 9
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L Joseph Callan @
Executive Director of Operations x US Nuclear Regulatory Commission /g'Tel O ~.5 L Washington, D C. 20555 ,
Dear Mr. Callan:
Please find enclosed a petition pursuant to Section 2.206 Title 10 of the Code c. Federal Regulations (CFR). The Prairie Island Coalition (PIC) hereby petitions the Nuciear Regulatory Commission (NRC) to suspend for cause the Northern States Power Co. (NSP) .
Materials License No SNM-2506 ceeded to operate an Independent Spent Fuel Storage Installation (ISFSI) at the Prairie Island Nuclear Generating Plant (PI). t A thorough review of the procedure developed by NSP for unloading Transnuclear dry storage casks (TN- 1 in use at PI is necessary at this time because it is apparent that conditions for safely unloaomg TN-40 catks after a storage period have not been established. By operating the ISFSI at P1 prior to establishing safe unloading conditions, NSP is violating the requirements of 10 CFR 72.122(1) and other rules and regulations of the United Statcs Nuclear Regulatory Commission. Toward this end, Petitioner also requests formal rulemaking proceedings under 5 U.S C. 553(e) to examine the issues addressed herein. .
T%nk you.
Sincerely, "r
George Crocker, Steering Committee Prairie Island Coalition {
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BEFORE THE' UNITED STATES NUCLEAR REGULATORY COMMISSION Docket 72-10 IN THE' MATTER OF:- )
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PRAIRIE ISLAND COALITION, ) , Dx Petitioner, ) x s
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-Respondent. )
'N Petition , ,
~ Pursuant to 10 CFR Part 2.206 of the Commission's regulations, .
'the Prairie Island Coalition petitions the Nuclear Regulatory '
Commission (NRC) to:
- 1. 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 $2.206 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 vioAated.10 CFR 572.122 (f) by using a cask design that requires periodic soal maintenance and emergency seal replacement that must be performed in the plant storage pQol. 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.
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- 3. Determine that NSP violated 10 CFR 572.122 (h) by using a
. 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 procedures adequate to safely unload and decommission the TN-40 casks.
- 5. Determine that NSP violated 10 CFR 572.130 by using the TN-40 cask and failing to make provisions capable of accomplishing the removal of radioactive wastes and contaminated materials at the time the ISFSI is permanently decommissioned. This failure may prevent decommissioning.
- 6. Determine that NSP violated 10 CFR 572.11 by failing to provide and include complete and accurate material information regar ;' ng maintenance and unloading of TN-40 casks in their ISt I application and in subsequent submissions regarding cask maintenance and unloading issues.
- 7. Determine that NSP violated 10 CFR 972.12 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 public review of maintenance, unloading, and decommissioning processes and procedures in question and an opportunity to comment on draft findings after investigation by the NRC.
- 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" I for air unloading of casks and transfer of the fuel.
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- 13. Under 5 U.S.C. 553(e), Petitioner requests a formal
.rulemaking proceeding to solicit information and review i 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 snendment of current licenses and '
rules for prospective licensing proceedings to require demonstration of a safe cask unloading ability before a cask may be used at an ISESI.
Preliminary Matters and Facts j
- 1. The Prairie Island Coalition (he'reinafter "PIC")
incorporates herein by reference the facts, argument, and !
conclusions of the Prairie Island Indian Community's 52.206 l Petition dated May 28, 1997.
- 2. PIC was established in 1990 for the purpose of location and
' dissemination of information regarding dry cask storage, andy opposition to NSP's plans to construct and operate an Independent Spent Fuel Storage Installation (hereigafter "ISFSI") at its Prairie Island Nuclear Generating Station (hereinaf ter "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 relat;.onship with the NRC regarding PI nuclear operations. PIC filed a S2.206 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 vc p*
l the NEC 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 l
issues, and~has met and exchanged written communications L with NRC staff about these issues.
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.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 long period of storage..." Exhibit A, February 25, 1997 Letter from NRC's Gail Marcus to George Crocker, Steering 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
! period of time.
- 6. Irradiated fuel in storcge casks' will experience thermal Thermal
! . shock when a cask is reflooded prior to unloading.
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 ,
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 coclant..." 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 proceduras Lapact worker or off-site radiation exposure.
. 7. Thermal shock may cause fuel assembly degradation. In the February 25, l?97 letter, 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 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 L
disintegrated fuel could be physically unloaded at all.
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- 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.
- 9. In a letter dated July 10, 1997, from Beth A. Wetzel of the NRC staff to NSP, Ms. Wetzel requests additional information regarding the PI spent fuel special ventilation technical 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 step further than Ms. Marcus in her letter (Ex. A),
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 pool ventilation system to adequately vent, contain, and 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
- 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 l
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.
l 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 terms. Exhibit E, April 16, 1997 NRC Memo from Jack Roe, Director, Division of Reactor Projects III/IV, Office of f Nuclear Reactor Regulation, to Cynthia D. Pederson, l 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 ible to immediately retrieve waste. See 10 C.F.R. S72.122 (1) . In his explanation of why licensee's ability to " someday, i somehow, maybe" retrieve spent fuel from storage would meet j the regulatory requirements, he fails to take into account
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I 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.
- 12. Mr. Roe also stated that:
[S]taff has not identified the unloading of a cask as a
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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 rationale for allowing a utility to operate where there is not enough room in the spent fuel l pool to unload immediately, i.e., at Prairie Island, or j 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 tr.at: ,
[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 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 walds.
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- 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 unacceptablo 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 unloaded.
- 15. Another reason the TN-40 ca.sks 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 (Minn. 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.Dccision and upon which all nuclear waste storage facilities are premised.
September 11, 1990, Waste Confidence Decision Revibw, 54 CFR ;
39767. Again, this is another scenario where the NRC's '
, j anticipation of the necessity of unloading is inadequate.
- 16. Yet another scenario where unloading is required is.for ]
decommissioning. NRC authority rests on the requirement i
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 l 1
s 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 issues that remain unaddressed, and NSP has not demonstrated 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 it could lead to future fuel handling and retrievability problems, is unacceptable. Ex. F, 4/15/97 NRC's Susan Frant Shankman Letter to Sierra Nuclear. In that particular case, the letter writer is concerned with degradation due to escape of helium, and emphasizes *that:
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. !
I Loss of helium from the TN-40 cask is an anticipat'ed event, ,
i 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. j 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. NSF 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. I Exhibit C, INEL Letter, October 18, 1990; Exhibit D, INEL l 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 r* ,
' unloaded. In the October 18, 1990 letter, the writer
. declared:
[T]he canisters had " setup"Lin 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 Fuel 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 othet six
. canisters in the TN-24P, it is possible, although not ,
- probable, that additional canisters may be unremovable, '-
it is also possible that canister. number 18 ig 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.
- 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
.placed in the spent fuel pool, and that replacement of the
[ seals is completed in the pool. Yet, as demonstrated by w Beth A. Wetzel's 7/10/97 Request, there are unresolved ,
safety considerations recognized by the NRC, primarily l ventilation of the flash steam produced by introduction of i 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.
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- 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.
NSP has also made this statement under oath in an Affidavit, andzin'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 _ tde 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 ?ower 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-Fuel 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 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. ,
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l NSP's claims that the TN-40 casks can be unloaded are
. baseless.
l 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 casks. NSP has not addressed or resolved this problem and has provided inaccurate and incomplete infonnation regarding this issue. .
NSP has violated 10 CFR 72.122(h) because the fuel is subject to degradation 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 infofmation regarding this issue.
NSP HAS violated 10 CFR 572.130 by using the TN-40 cask and failing to make provisions that facilitate the removal of radioactive wastes and contaminated materials at the time the ISFSI is permanently decommissioned. This may prevent decommissioning in so far as a TN-40 cask that cannot be unloaded can therefore not be decommissioned.
NSP has violated 10 CFR 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. 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.
i NSP has violated 10 CFR 672.12 by deliberately and knowingly i submitting incomplete and inaccurate material information regarding maintenance and unloading of TN-40 casks in their ISFSI application and in subsequent submissions regarding cask s-
- e -r 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 '
submission 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 1 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 rogulate 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 evaluation 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: 9&f !
F em
(.f' kj UNITED STATES NUCLEAR REEULATORY COMMISS ON Exhibit A' j
WASMH0STON. C.C. EEEME01 i
/ February 25, 1997 i
George Crocker, Steering Comeittee Prairie Island Coalition P.O. Box 174 Lake Elmo, NN 55042 6
Dear Mr. Crocker:
As the lead manager for dry cask issues in the Office of Nucleav oeactor Regulation (NRR), Nuclear Regulatory Commission (NRC), I as regonding 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-res 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 - 1 i
limits for maximum radiation exposures to plant personnel are defined in Parti 20 of. Title 10 of the Code of Federal Reaulations (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 expected 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 accident involving a dry storage cask at Prairie j
!sland would result in less exposure to the general public than does this j 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 perfomed 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 under various conditions, and performing various activities associated with reactor facilities. In the absence of actual exoerience in unloading spent fuel from a cask following a long period J
, , d G. Crocker of storace. a annaral understandina of technical canabilities and related experiences enablos the NRC staff to assess the adecuacy of a licensee's procedures for un' oadina dry storace casks. For those examples of cask unloadings 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 to 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 license. >
imloading procedure (Enclosure 1) requires personnel to determine if additional steps or precautions are warranted based on the analysis of the 9, sample from the cask cavity. Additional surveys and samples are taken
~throughout the unloading process to ensure that the radiation doses received by licenses 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 coolant thqt may occur i if a cask was unloaded after a period of storage, engineering evaluations andy ,
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 1
- during storage, the unloadina may reauire additional filterina and even vacuumina debris from the bottom of the cask. Such steps would be developed 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 tr retrieve fuel pellets, as a result of several cases of fuel assembly damage that occurred during reactor operation. Although licensees would be able tr develop means to retrieve degraded fuel assemblies from a dry storage cask, ;
the accumulated occupational dose to perform this activity may be increased l from the previously mentioned estisaates. Fuel reactivity for criticald tv considerations could increase only under very idealistic and hich' y un'ikel, iisintearation oatterns in the fuel, toon detection that fuel disintecrati;u
- ind occurred. special measures would be deve'oced and isolemented to assure 6 adequate safety margin is maintained during unloading.
Some SARs do state that unloading is basically the reverse of loading and tU statement, in a general' sense, is true. However, such statements may tend o 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 olan for dry cask storaae and related statementi made by the NRC staff. includina those by Mr. Kualer. the staff was !
emphasizine that icensees need to identify the conditions and complicatie I
gr . ,.
G. Crocker .
that are associated with the unloadina crocess and ensure that unloadina 3rocedures address those concerns. The unloading procedure for the dry storage casks at Prairie Island was inspected by the NRC staff and, following minor revisions, was found to provide adequate guidance to control the unloading 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 infomation addresses your concerns. Please contact William Reckley on 301-415-1314 if you have any additional questions or ,
concerns.
s
' Sincerely, JaML-GailH.Marcus,ProjectI)irector 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, MN 55069 cc w/o enc 1: see attached page t ,
- - -.vg\ UNITED STATES
'N Exhibit B NUCLEAR REEULATORY CBMMISSION 5 c ,,,,,,,, g,,, ,,,,,
,,,,, July 10, 1997 i
Mr. Roger 0. Anderson, Director (
l Licensing and Management lasues -
Northern States Power Company 414 Nicollet Mall -
3 Mnneapolis, Minnesota 55401
SUBJECT:
RE6UEST FOR ADDmONAL INFORMATION ON THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2, '
AMENOMENT OF SPENT FUEL POOL SPECIAL VENTILATION TECHNICAL SPECIFICATION (TAC NOS. M98782 AND M98753)
Deer Mr. Anderson:
By letters dated May 7,1997, and supplemented May 30,1997, Northom States Power Company submitted a recuset to amemt the Prairie faland Technical Specifications pertaining to the spent fust pool special ventRetion system. In order to review the proposed changes the staff requires some additional information. ~ Our request for additional information (RAI) is enclosed, in order to continue our review of your submittal on an expedited basis, please '
provide your response to the staff's RAI as soon as proctical if you have any <
questions regarding the content of the RAI, please contact me et (301) 415-1355.
r Sincerely, v h.
Both A. Wetzel, Pr Manager f Project Directorate lil-l ;
Division of Reactor Projects BilAV l Office of Nuclear Reactor Regulation l Docket Nos. 50-282. 50 306
Enclosure:
As stated i
oc w/enci: See next page l l
i I
l l
I
f* ,a 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 I
be filled with water. The caution prior to step 8.27 reeds, "The water / steam mixture from the vent port hose may contain some radioactive Oas. The ares directly above where the hose la discharging shall be closely ' -
monitored to determine if Gwe is a radiological hazard." is the spent fuel pool special ventilation syr.hm operable during the performance of this step of the unloading procedurd/ if the spent fuel pool special ventilation system is inoperable during this step and other portions of the unloading procedure because the overhead crane is supporting the cask through the open spsnt 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 to CFR Part 20 and Part 100 requirements are not exceeded.
- 2. Section 5.5 of the Prairie Island ISFSI [ independent spent fuel shwage ,
installationi safety analysis report ISAR) states in part, "After moving the cask into the fuel pool ares, the cavity will be depressurized and,the cask lowered into the spent fuel pool." However, Step 8.4 of procedure D95.2 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 depressuriastion? Does the SAR assume that the spent fuel pool special ventilation system will be operable during the cask depressurisation evolution?
- 3. When the spent fuel cask is filled with water prior to unloading the fuel (per Step 8.27 of D95.2, "TN 40 Cask Unloading Procedure"), discuss the likelihood that this will result in cracking of the spent fuel rods due to the !
interaction of the cool spent fuel pool water with the hot fuel alaments. If any fuel cracking la predicted, list the expected redlonuotidos and quantities that will be rolessed into the cask and into the fuel building when the cask is vented. If the filtered ventilation system is not operating during cask i venting, describe how you plan to detect and prevent these radioactive gases from being released into the enconment.
1
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I' ** l hsanceq g k UNITED STATES 4 ?> NUCLEAR REGULATORY COMMISSION U WASHINGTON. D.C. segeM001
.....$ April 16. 1997 MEMORANDUM TO: Cynthia D. Pederson, Director Division of Nuclear Materials Safety Region III
, L-14 --
FRON: Jack W. Roe Ofreciio~r - Division of eactor Projects III/IV Office of N 1 ear Reactor Regulation
~
SUBJECT:
TASK INTERFACE AGREEMENT 96-0440; DEFINING DRY CASK STOPAGI TERMS (TACN05.N97346ANDM97347) In response to your request dated November 26, 1996, NRR/DRPW and NMSS/SFPC have discussed the questions raised and offer the following clarifications regarding the , terms ready retrieval and structural defects. The two basic reasons to return a cid to the spent fuel pool and unload the spent fuel assemblies are either to (1) ratrieve the fuel assemblies for further processing or disnosal, or (2) respond to an event or condition that has potentially degraded the design requirements estabitshed for the cask. The staff has not identifist the unloading of a cask as a renoiced protective __ _measurolo_ be taxen within a specified time in order to limit the offsite consequeces of an accident involvino the release of radioac'tive materiai from ' a storage cask. - In regard to the CAQpi"""" unt cask designs must allow retrieval of the spent fuel for further precessino or disposal (10.ctM Iz.1zz(1)), the nac has consistently taxen a por.!tton that 11censees can satisfy this requirement without maintaining the capability to retrieve tne spans fusi i v. a cask wTthin a specirieo perioo or time, ano may, ir necessary, oeveiop aiternate up6 Uns for fuel retrieva! if a cask unloaolng cannub os i wiaisiy suppuri,ed _ duw w a snortaae or soace In a spent ruei pooi. inis is constaerec acceptable because licentues have a great deal of flexibility in their ability to schedule and plan for tho transfer of spent fuel from a storage cask to another cask for storage or shipment. Several of the actions req: ired by ISFSI technical specifications or cask certificates of compliance spacify that, in the case of certain events or - l conditions, a cask may need to be~ unloaded, or otherwise returned to a safe i storage condition. The NRC staff has stated that the potential need to unload a cask in response to an event or condition in the technical specifications or certificates of compliance does not require licensees to maintain a continuous ability to unload a cask within a specified time. This position is based on (the absence of an identified event or condition involving the storage casks gthat would result in an immediate threat to public health and safety. The position is reflected in past NRC decisions such as the acceptability of (1) licensees not having to maintain space in spent fuel pools to acconnodate CONTACT: William Reckley, NRR (301) 415-1314 APR 2t W.
C. Pederson unloading 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 July 1993, stated that its review of the accident analyses determined that, " Dose equivalent consequences, from a single cask, to any individual, from direct or indirect radiation.and gaseous activity .' release after mstulated accident events,are less than the 50 mSv (5 res) dditionally, in its Environmental ' limit establisaed in 10 CFR 72.106 b)." A Assessment, dated July 28, 1992, th(e staff assessed the accident dose at the site boundary as, "... a small fraction ... of the criteria specified...." and found that, "The doses are also much less than the Protective Action Guides . established by the Environmental Protection Agency (EPA) for individuals l exposed to radiation as a result of accidents." 8ecause 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. However, following certain events or conditions, the licensee is required to take corrective actions to ensure safe storage conditions and to perform 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 act have to be unloaded inusediately to maintL*n . safe storage conditions. The licensee would have time to consider available , options, 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 unloading. Operational safety problems are those that involve gross rupture of the fuel cladding such that significant quantities of fuel material and fission products are released to the 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) and 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.
C. Pederson . definition for " gross cladding defect" hat M n incorporated into NUREG-1536,
" Standard Review Plan for Dry Cask Storage systems," which was recently issued in final form.
In the specific case of Prairie Island, neither 10 CFR 72.122(T') or specific ISFSI technical specifications introduce additional requirements for the fuel handling equipment used to actually load or unload the fuel assemblies into the cast since such matte regulationsandlicenses.psareaddressedunderexisting10CFRPart50 The structural requirements defined by the ISFSI ' sAR and technical specification are satisfied e'/en if it is necessary to use a special handling tool to overcome problems in lifting 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 saalyses 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.,1ons If the licensee's in 10 CFR Part 50 actions areand the 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 specification 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 specfal fuel handling procedure increases the probability of a fuel handling accident - within the reactor facility, actions or inquiries from the NRC staff should be presented in the context of regulations such as Appendix B to 10 CFR,50 or 10 CFR 50.5g. 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. Hehl, RI
- 8. Mallett, RII R. Scarano, RIV 8
Prairie Island ISFSI Technical Specification 1.3.2, " Fuel and Cask Handling Activities," states: Fuel and cask movement and handling activities which are to be performed in the Prairie Island Nuclear Genersting Plant Auxiliary Building will be governed by the requirements of the Prairie Island Nuclear Generating Plant Facility Operating Licenses OPR-42 and DPR-60 and associated techntral specifications. l
J NI(C Inspecti::n Report - Sierra Nuclear Corp. Selartad Ranarts Index l News and Information l NRC Horne Page l E-mad Aprd 15,1997 Mr. Art J. McSherry President Sierra Nuclear Corporation - One Victor Square Scotts Valley, CA 95066
$UBJECT: NRC INSPECTION REPORT NO. 72-1007/97-204 AND NOTICE OF NONCONFORMANCE
Dear Mr. McSherry:
This letter refers to the inspection conducted March 17-21,1997, at your facility in Scotts Valley, California, and at two ofyour fabrication contractors' facilities: March Metalfab, Inc., in Hayward, California; and Nrr-Cal Metal Fabricators, in Oaldand, California. The term avarnined information about. seal weld failures on dry spent fuel storage casks at the Palisades and Arkansas Nuclear One (ANO) nuclear power plants. Additionally, the team assessed the adequacy of your corrective actions taken for the findmgs identified in, . Inspection Reports 72-1007G6-204 and 96-208, regarding the Model VSC-24 dry spent fuel storage system manufactured under Certi6cate 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 theinspectir team found that you failed to meet certain Nuclear Ragni..a,y Commission l requirements. The tean. ntified four nonconformances regarding failures to perform work in accordance with your Quality /.ssuranse Program. The nonconformances were failures to (1) examine the potential
. generic aspeas of the shield-lid weld failures at ANO and Palisades, (2) submit a change to the Certi6cate of <
Compliance to correct a normnservative requirement for the drain-down time limit for a loaded cask, (3) submit a Safety Analysis Report cht .ge to correct the 1986 American Society of Mechanical Engmeers Code omission of nondestructive examination requirements for temporary attachments, and (4) control measuring j Mest equipment. . Two of the nonconformances raise safety concerns. First, the shield lid weld failures affect the 'mtegrity of a cask ee8a-aaa' boundary. The root-cause of the shield-lid failures and the potential for delayed cracking on loaded casks must be understood. Ahhough 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 hamilina 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 w->ptable. Second, the nonconservative Technical Specification for cask drain-down time affects the margin to criticality. N
rdgC.
, Sierya Nuclear Corporation's lack of timely and comprehensive action, in dealing with these imponant safety issues, is a significant regulr_ tory concem. As the certificate holder, Sierra Nuclear Corporatirn is responsibla 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 meeting you j should be prepared to discuss your shon term and longer term corrective actions to address the issues and l concerns raised by our ia*Wion.
l 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 Nonconfonnance (Enclosure 2). We will consider extending the response time if you 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,
/ signed /
Susan Frant Shankman, Chief Transponation Safety and Inspection Branch Spent Fuel Project Office, NMSS -
Enclosures:
3
, s
- 1. Inspection Report 72-1007/97-204
- 2. Notice of Nonconformance Docket No. 72-1007 l
i
4 / Cobfirmetsry Actica Letter - ArkanSnS Nuclear - Selected Reports Index
- l News and Information l NRC Home Page l E-mail May 16,1997 CAL No. 97-7-002 Mr. C. Randy Machinnan
- Vice President, Opersoons ANO Entergy Operations,Inc.
1448 S. R. 333 Russellydle, AR 72301
SUBJECT:
CONFIRMATORY ACTION LE' ITER ,
Dear Mr. B+Masaa:
During the week of March 17,1997, U.S. Nuclear Relatary Commission staffiaW Skira Nuclea Ccrporation (SNC) and two ofits fabrication contractor facilities. SNC holds Certdicate of Com 1007 for the VSC-24 dry storage cask. This inspection focused on welding problems with VSC-24 cas used at the Palisades and Arkansas Nuclear One (ANO) nuclear power plants. The problems w weldsjoining the cask shield lid to the multi-assembly storage basket (MSB). The Palisades w occurred in March 1995 and the first instance at ANO in December 1996. AAer the record SNC inspecti problems arose while welding another ANO cask on March 26,1997. NRC is concerned ab egeguntered with the welds joining the shield lid to the MSB, since this weld
~
~ '
b is part.of bl the c+d=
, boundary of the VSCWNrthermore the weld between the MSB and the structurallid
, may e far weld probleris suscepu e may to the same failbre_mechsnisms.gs_the shilel lid weld. It is possible that these partie d faiwes '
wugh'such not develon uritil after cask welds have underaone non datructive examiRatiKn'~ Alt would not pose an off-site threat to public health and safety, such an occurrence would caus _ / [ad'mg helium _ p'roblems,. and retnevabahty atmosphere inside the MSB. This conditio The March 1997 inspection revealed that neither SNC nor the user licensees had performed root-cause analysis of the first two weld problems. An understandmg of the root cause is essential to Preventing recurrence when weldmg future caska, and to assessing the possibihty of additiona problems, perhaps undetected or delayed, in loaded casks. On May 6,1997, NRC held a SNC ..ri:: Nadves to discuss SNC's implemented and planned actions in response to the weld pro inspection findings. Rg=:=dves of your staff attended this meeting. As stated at this meet remains concerned that the root cause(s) of the weld problems have not been conclusively determine Puriuant to a May 14,1997, telephone conversation between Randy Edington t.nd Charles Haugh Deputy Director of the Spent Fuel Project Office, Of5ce of Nuclear Material Safety and S understanding that you will take the following actions before loading additional VSC-24 casks wit nuclear fuel: l l l l
(1) Determine that your welding and inspection pracuces p evide reasonable assu mce tnat crusung, including possible undetected c: delaved cradang : vill not ccur to the welds sea ting the shield lid a structurallid to the MSB If necessarv. modify your weleg p'ucesses to inhist recurrence cf these weld problems. iC-24 cask with spent fuel, f (2) On completion of tius actirm, and m.len .M w tdm fairy m&:' you will submit to the Directot. Ofhe af he rt .4beruo Safesy and bw ds, .s written description any procedural or design modi 6catirsenade with u:ye % .ittm .' ne.suhtts thould include the technicaljustification f:a cach modifaate.n htqud1hs:ssai secuW w.sem to William F. Kane, Director, Spent Fuel Projea ()f5ce, and to yow te ghms Adminimmt 0:4s may include in this respon the information required by item 2.bekiw, confetuopf, ae act6d cequs.d ayitem 1 above.)
)
Pursuant to Section 182 of the Atom >cEnergy.Act,42 U.S.C 7232,;m ne required to: (1) Notify me immediately if your understanding differs from that set fcnh above; (2) Notify me in writing when you have completed the actions addressed in this Confirmatory Act l Issuance of this Confirmatory Action Letter does not precludq issuance of an order formalmng the above commitments or requiring other licensee actions; nor does it preclude NRC from taking enforcement a f;r violations of NRC requiremems that may have prompted the issuance of this letter. In addition, fai)'tre 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 f will be placed in the NRC Public Document Room (PDR). To the extent possible, your respons include any personal privacy, proprietary, or safeguards information so that it can be placed in the l without redaction. However, if you find it necessary to include such information, you should clearly indic the specific information that youdesire not to be placed in the PDR, and provide the legal basis to your request for withholding the information from the public. Sincerely, j Malcolm R. Knapp, Deputy Director ' OfHee of Nuclear Material Safety and Safeguards Dockets 72-1007, 72-13, 50-313, 50-368 i l l
' ISFSI SAR TABLE 5.1-2 ANTICIPATED TIME AND PERSONNEL REQUIREMENTS FOR CASK HANDLINC OPERATIONS No. of Tine Avg. Distance Operation Personnel M (ft) from Cask Raceivina
- 1. Unloading (A1) * * * ,
- 2. Inspection (A2 through A7)
- 3. Transfer to cask * *
- loading pool (AS) e--k r,-av.,p 1
- 4. Imwer cask into pool (51) * * *
- 5. Imad fuel (B2 through 54) 5 5
- 6. Place lid on cask (55) 30 5
- 7. Lift cask to pool surface (56) 5 3 5 120
- 8. Install lid bolts (86) lb. Tre f c contani co 3 60 10 area (812) .
. i I
Decone==ination Area t 3 120 3
- 11. Decontaminate cask (C1, C2) 5 2 30
- 12. Remove vent plugs
- 13. Drying, evacuating, 480 5 backfilling (C3 through C13) 2 3 15
- 14. Install top neutron shield C14) 2 -
- 15. Install pressure .
30 5 transducers (C15 through c17) 2 *
- Pressurize interspace (C18) *
- 16. 30 5
- 17. Check leakage (C19) 2 30 5
- 18. Check surface temperature (C20) 2 3 2 30
- 19. Check surface dose rate (C21) 30 5
- 20. Install protective cover (C22) 2 60 5
- 21. Lead on transport vehicle (C23) 3 60 10
- 22. Transfer to storage area C24) 3 TABLE 5.1-2 m.2 W I
r ISFSI SAR TABLE 5.1-2 (Continued) ANTICIPATED TIME AND PERSONNEL REQUIDEMENTS FOR l CASK HANDLING OPERATIONS f Doeration No. of Time Avg. Distance Personnel ,(51D.). (ft) from Cask l t Storaae Area
- 23. Unicad from vehicle .
position in location 60 5 (D1, D2, D3) 5 30 3 l 24. Check surface dose rate (D6) 5
- 25. Connect pressure instrumentation (D4, D5) 5 30 5 Periodie Maintenance 1
15 5
- 1. Visual surveillance (NA) 2
- 2. Repair surface defects (NA) 2 60 3
- 3. Instrument testing and
- 180 5 calibration (NA) 2 3
4 Instrument repair (NA) 2 60 .
- 4
/
' 3 1950** 8 0 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 e replace lid seals, and return cask to ISTSI pad.
l TABLE 5.1-2 REV. 2 9/91
's ' ISFSI SAR TABLE 5.1 2 (Continued)
ANTICIPATED TIME AND PERSONNEL REQUIRDGNTS FOR CASK HANDLING OPERATIONS No. of Time Avg. Distance Ooeration (ft) from Cask Personnel fain), Storane Area
- 23. Uniosd from vehicle '
position in location 5 ( (D1, D2, D3) 5 60 5 30 3
- 24. Check surface dose race (D6)
- 25. Connect pressure instrumentation (D4, D5) 5 30 5 l
Periodic Maintenanes 15 5
- 1. Visual surveillance (NA) 2 3
- 2. Repair surface defects (NA) 2 60
- 3. Instrument testing and 180 5 calibration (NA) 2 3
Instnament repair (NA) 2 60 .
- 4. .
/ Maior Maintenance (once in 20 years) ,
1950** 8
- 1. Replace cask lid seals 3
- No measurable dose associated with this activity. Enerefore, 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 ISFSI pad.
l TABLE 5.1-2 REV. 2 9/91 1 1
M" UNITED STATES
- f *%k2 3
NUCLEAR REGULATORY COMuiSSION as -
*n ! soiwAnnemuinoAo USLE. ILUNOIS SDM-4361
*e..* - June 30, 1995 Mr. E. Watzl, Vice President Nuclear Generation Northern States Power Company 414 Nicollet Mall '
Minneapolis,PW 55401 .
Dear Mr. Watz1:
This refers to the special NRC fnspection 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 Safe ~ty and Safeguards. The purpose of this inspectkn was to evaluate the acceptability of the as-built TN-40 cask and to assess your performance relative to dry cask storage including the nreooerational 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 fres the requirements of 10 CFR Part 72.82(a) 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. I On May'12 you loaded the first cask with spent fuel. The enclosed copy of our inspection report identifies areas examined during the inspe': tion. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and
' inrtervie'ss with personnel. .
Based or, the results of this inspection, we concluded that you were ready to safely load spent fuel into the TN-40 dry storage cask and transport this cask to tia onsite ISFSI. We also did not identify any safety concerns with the ; subject cask. However, one violation of NRC requirements was identified l during the course of this inspection, as specified in the enclosed Notice of Violation (Notice). This violation pertained to cask handling, loading, and unloading 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 your receipt of this Notice, a written statement of explanation, we note that l this violation had been corrected and those actions were reviewed.during this inspection. Therefore, no respense with respect to this violation is j required. However, we are disappointed that NRC inspectors, rather than your cwn staff, identified these procedural deficiencies. 1
e- o E. Watz1 We also identified several weaknessu nith your mric perfursomca rclative li poor oversight to dry cask storage activitiu.. Tearse 'enkarwn 'inclace* of vendor activities until hte in the dra rust starage pro.ywct, 2) lack of effective engineering involvenerd is vesht -fabricat>m xtivities; 3) the ineffectiveness of your quality assaraece organhation in asa.ssing vendor performance during the cask fabrication prscess; 4) the ahsence of a comprehensive plan for inspecting, auditiseg, and moeftering dry cask storage activities onsite, particularly those activitles associated with the 10 CFR Part 50 license; and 5) overall poor planning for dry cask storage activities. '
/,
Based on the above weaknesses and as discussed at the exit meeting on April 28, we request that you provide us with a formal performance improvement plan documenting the specific corrective actions you have already taken and . .. those activities. you plan to implement to address the above weaknesses ir. d inspection report. We will continue to evaluate the effectiveness of your corrective actions to improve your perfumance in dry cask activities during future NRC inspections. ,
!n accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, the enclosure, and your response to this letter will be placed in ,
the NRC Public Document Room.
.The response requested by this letter is not subject to the clearance r procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.
We will gladly discuss any questions you have concerning this inspection. Sincerely, fM&. k m=& Edward G. Greenaan Senior Oversight Manager Region III Dry Cask Activities Docket No. 50-282 Docket No.'50-306 Docket No. 72-10
Enclosures:
- 1. Notice of Violation
- 2. Inspection Report No. 50-282/95002; 306/95002; 72-10/95002(DRP)
See Attached Distribution
NOTICE 0F VIOLATION Northern' States Power Company Dockets No. '50482; 50-306; 72-10 Prairie Island Nuclear Plant- Licenses No. DPR-42;-DPR-60; SNM-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 licensee to establish, maintain, and execute e'
-a quality assurance (QA) program with regard to an Independent Spent Fuel Storage Installation (ISFSI) that satisfies each of the applicable criteria of Subpart G, " Quality Assurance.' In meeting the Part 72.1429) requirement, 10 CFR Part 72.142(d) accepts a Commission-approved quality assurance program *
- which satisfies the applicable criteria of Appendix B to 10 CFR Part 50. As such, the ISFSI Safety Analysis Resort ' states that the previousiv *aaroved Northern States Power QA program wsich satisfie's applicable criteria of 10 CFR Part 50, Appendix B,'will ne applied to activities, structures, systems, and -
components of the ISF51 commensurate 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. trary to the above, cask handlina. loadina. and unlandian activities were at meatmnar nv annrovet )roceduret af a tvns ADDropriate to the
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Wcumstance'slas evidencec )y the(rollowing examples: j
- 1. . . Surveillance Procedure. SP 1077, "Special Lift Fixture for the TN-40 Cast," did not address dimensional checks of the special lifting device, as required. -
- 2. Surveillance Proceuure, SP 1075, "TN-40 Fuel Selection and Identification,' did not incorporate 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 occumented."
- 3. Procedure D95.1, "TN-40 Cask Loading Procedure," s prerequisites section that SP 1077 be performe@pecified ays prior to loadingin the ,
a cask. However, the TS 4.19-requirement to per T om a visual 4==aetion j of the lifting device (lift beam and extension) ard verify operability) ; of the device 7 days or to use, was not identiftea m un.a. murs also was no procedure identifying actions required to verify operability of the lifting device. i l
, Notice cf Violatten '4. Procedure 095.1, "TN-40 Cask loadina Procedure
- did not include a step to perform radiation mrvave of the cask surface before moving a cask to tne 1stsl, as reautred by TS 4.6.1.
- 5. Procedure D95.2, "TN-40 Cask Unloading Procedure,' did wot adeauntely address the TS-requirement to <=nie tne spent ruei pool for boron concen6rauon wanin tour hours of flooding the cast cavity for '
unloading tne ruei assemonies. . .
- 6. Procedure D95.2, "TN-40 Cask Unloadino Procedule" did not contain a hold point to ensure worn wousa not continue unt 1 The results of the Inner e cask volume sample had been reviewed. Ints procacural nold point is impor6an6 6e ensure snas an unpianned and unmonitored release path is -
not created while the cask is in the Auxiliary Building. , i
- 7. The licensee did not have a procedure for conducting 10 CFR Part 72.48' safety evaluations.
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This is a Severity Level iV Violation (Supplement I) (50-282/95002-0'1; 50-306/95002-01; 72-10/95002-01(DRP)). With respect to this violation, the inspection showed that steps had been tak:n to correct the identified violation and to prevent recurrence. Crnsequently, nc 'eply to the violation is required and we have no further , questions regarding this matter. r Dated at Lisle, Illinois this 30th day of June 1995
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c . ., ., - o While the inspectors recognized that finalizing the loading and l 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 perore tne dry Ar ror cumpie, technicas apecirication requirements were not effectively incorporated into the loading and unloading procedures not como:ete rev<ew and I (paragraph 3.2). In addition, the licennes die _ approval of the unloading proceoure unti the c av Fo1owngsube'ssionf thin renart imonied or rne creoperational test report. Su ,wntssion of oad a cask with spent fuel and f that the l' censee was ready to subsequentiy unload the cask, if necessary. O The licensee did not take a disciplined approach to inspecting the fuel . designated for cask storage as evidenced by weaknesses identified by the inspectors during observation of fuel inspection activities (paragr,aph
- 7.3).
o Some weaknesses were noted with the licensee's documented basis for safety evaluation conclusions (paragraph 8.2). V 3
$p; rational ch:cks of vehicle brakes, lifting eouipment, turnetbles, jacks, and cask links.-
3.1.5 Surveillance Procedure. SP 1075. "TN-40 Fuel Selection and 5 Identification"
)
The inspectors reviewed SP 1075 and the cask loading procedun, D95.1, to verify that selected Technical Specification (TS) requirements had beenSur incorporated into procedures. fu31 assemblies which satisfy the criteria of TS 3.1.1 would be loaded into the cask, are defined in TS 4.I. TS 3.1.l(6) required that, " fuel assemblies known or suspected to have structural defects or gross cladding failures (other than pinhole leaks)ility sufficiently severe to adversely affect fuel handling Theand transfer licensee capab - originally - shall not be loaded into the cask for storage." htended te visually inspect fuel assemblies designated for loading with binoculars 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 III staff on fuel insoection techniques, the licensee elected to use video recording The inspectors considered tt.!s a equipment to perform the fuel inspection. preferable TS 3.1.1. method for identifying fuel anomalies and ensuring 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 i not incorporate the requirement of TS 4.1.2, which stated that "before l inserting a spent fuel assembly into a cask... the identityThe of each inspectors fuel 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 Based on observations of the verification of fuel assembly identification. actual fuel inspection, the inspectors concluded that the licensee met all TS j The failure to incorporate the requirements for fuel identification. requirements of TS 4.1.2 into SP 1075 is considered ar Sxample of a violation of CriteMon Y 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 Unloadir.: Procedures l The inspectors reviewed the ic,ading (D95.1) and unloading (D95.2) procedures for technical adequacy and to determine if the lessons learned from the preoperational testing / dry ra had been appropriately incorporated into the procedures. 3.2.1 D95.1. "TM-40 Cask loadina Procedure" The original 095.1 procedure specified in the prerequisites section thatHowever, the T SP 1077 be performed 30 days prior to loading a cask. Specification (TS) 4.19 requirement to perform a visual inspection of the 10 ! I}}